SEC Petition Evaluation Report 

Petition SEC-00093 

Report Rev #: 1_ Report Submittal Date: April 28, 2009

Subject Expert(s): Ray Clark, Gene Potter 

Site Expert(s): NA 

Petition Administrative Summary 

Petition Under Evaluation 

Petition # Petition 

Type 

Petition B 

Qualification Date 

DOE/AWE Facility Name 

SEC-00093 83.13 October 19, 2007 Santa Susana Field Laboratory-Area IV 

Petitioner Class Definition 

All employees who worked in all areas of Santa Susana Field Laboratory-Area IV during the time period from 1955 to 

the present (which incorporated the post-1987 remediation period). 

Class Evaluated by NIOSH 

All employees of the Department of Energy (DOE), its predecessor agencies, and DOE contractors and subcontractors 

who worked at Area IV of the Santa Susana Field Laboratory from January 1, 1955 through December 31, 1965. 

NIOSH-Proposed Class(es) to be Added to the SEC

All employees of the Department of Energy (DOE), its predecessor agencies, and DOE contractors and subcontractors 

who worked in any area of Area IV of the Santa Susana Field Laboratory for a number of work days aggregating at least 

250 work days from January 1, 1955 through December 31, 1958, or in combination with work days within the 

parameters established for one or more other classes of employees in the SEC.

Related Petition Summary Information 

SEC Petition Tracking #(s) Petition Type DOE/AWE Facility Name Petition Status 

N/A N/A N/A N/A 

Related Evaluation Report Information 

Report Title DOE/AWE Facility Name 

N/A N/A 

ORAU Lead Technical Evaluator: Ray Clark ORAU Review Completed By: Daniel Stempfley 

Peer Review Completed By: [Signature on file] 4/28/2009 

Lara Hughes Date 

SEC Petition Evaluation Reviewed By: [Signature on file] 4/28/2009 

J. W. Neton Date 

SEC Evaluation Approved By: [Signature on file] 4/28/2009 

Dave Sundin for Larry Elliott Date 
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Evaluation Report Summary: SEC-00093, Santa Susana Field 

Laboratory-Area IV 

This evaluation report by the National Institute for Occupational Safety and Health (NIOSH) 

addresses a class of employees proposed for addition to the Special Exposure Cohort (SEC) per the 

Energy Employees Occupational Illness Compensation Program Act of 2000, as amended, 42 U.S.C. 

§ 7384 et seq. (EEOICPA) and 42 C.F.R. pt. 83, Procedures for Designating Classes of Employees as 

Members of the Special Exposure Cohort under the Energy Employees Occupational Illness 

Compensation Program Act of 2000. 

Petitioner-Requested Class Definition

Petition SEC-00093, qualified on October 19, 2007, requested that NIOSH consider the following 

class: All employees who worked in all areas of Santa Susana Field Laboratory-Area IV during the 

time period from 1955 to the present (which incorporated the post-1987 remediation period). 

Class Evaluated by NIOSH

Based on its preliminary research, NIOSH reduced the petitioner-requested class to a class that 

includes the time period from January 1, 1955 through December 31, 1965. NIOSH evaluated the 

following class: All employees of the Department of Energy (DOE), its predecessor agencies, and 

DOE contractors and subcontractors who worked at Area IV of the Santa Susana Field Laboratory 

from January 1, 1955 through December 31, 1965. 

NIOSH-Proposed Class(es) to be Added to the SEC

Based on its full research, NIOSH modified the petitioner-requested class to define a single class of 

employees for which NIOSH cannot estimate radiation doses with sufficient accuracy. The revised 

NIOSH- proposed class includes all employees of the Department of Energy (DOE), its predecessor 

agencies, and DOE contractors and subcontractors who worked in any area of Area IV of the Santa 

Susana Field Laboratory for a number of work days aggregating at least 250 work days from January 

1, 1955 through December 31, 1958, or in combination with work days within the parameters 

established for one or more other classes of employees in the SEC.

NIOSH cannot estimate internal exposures with sufficient accuracy during the period from 1955 

through 1958 (which is a period with limited internal monitoring data). This includes the time from 

the beginning of Area IV radiological activities in 1955, to late 1958 (selected as December 31, 1958 

for SEC evaluation purposes). This timeframe corresponds with the date after which an established 

bioassay program existed at SSFL and after which sufficient internal monitoring has been identified. 

With the exception of the class above, NIOSH has obtained sufficient information to support 

bounding doses for the entire evaluation period. 

In its original evaluation report (Rev. 0) for SEC00093, the NIOSH proposed class definition to 

included all employees of the Department of Energy (DOE), its predecessor agencies, and DOE 

contractors and subcontractors who were monitored while working in any area of Area IV of the 

Santa Susana Field Laboratory for a number of work days aggregating at least 250 work days from 
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January 1, 1955 through December 31, 1958, or in combination with work days within the parameters 

established for one or more other classes of employees in the SEC.

 

The class definition proposed by NIOSH in revision 0 was based on the infeasibility of reconstructing 

internal doses for the period from January 1, 1955 to December 31, 1958. Information available at the 

time Revision 0 of the report was completed indicated that all workers with the potential for internal 

exposures would have been monitored with a radiation dosimeter badge. The change in proposed 

class definition was triggered by additional information collected by NIOSH indicating that there was 

a possibility that not all workers who entered controlled areas were necessarily monitored for external 

radiation exposure. Because of this, NIOSH can not definitively rule out the possibility that some 

unmonitored workers could have incurred internal exposures. 

 

Feasibility of Dose Reconstruction

NIOSH cannot estimate internal exposures with sufficient accuracy during the period from 1955 

through 1958 (which is a period with limited internal monitoring data). This includes the time from 

the beginning of Area IV radiological activities in 1955, to late 1958 (selected as December 31, 1958 

for SEC evaluation purposes). This timeframe corresponds with the date after which a routine 

bioassay program existed at SSFL and after which sufficient internal monitoring has been identified. 

With the exception of this class, per EEOICPA and 42 C.F.R. § 83.13(c)(1) NIOSH has established 

that it has access to sufficient information to: (1) estimate the maximum radiation dose, for every type 

of cancer for which radiation doses are reconstructed, that could have been incurred in plausible 

circumstances by any member of the class; or (2) estimate radiation doses of members of the class 

more precisely than an estimate of maximum dose. Information available from the site profile and 

additional resources is sufficient to document or estimate the maximum internal and external potential 

exposure to members of the evaluated class under plausible circumstances during the specified period 

from 1959-1965. 

Health Endangerment Determination

Per EEOICPA and 42 C.F.R. § 83.13(c)(3), a health endangerment determination is required because 

NIOSH has determined that it lacks sufficient internal monitoring data for the period from January 1, 

1955 through December 31, 1958, which prevents evaluating internal dose with sufficient accuracy 

for the members of the proposed class during this period. 

NIOSH did not identify any evidence supplied by the petitioner or from other resources that would 

establish that members of the proposed worker class were exposed to radiation during a discrete 

incident likely to have involved levels of exposure similarly high to those occurring during nuclear 

criticality incidents. However, evidence indicates that some workers in the proposed class may have 

accumulated substantial chronic exposures through episodic intakes of radionuclides, combined with 

external exposures to gamma, beta, and neutron radiation. Consequently, NIOSH has determined that 

health was endangered for those workers covered by this evaluation who were employed for a number 

of work days aggregating at least 250 work days within the parameters established for this class or in 

combination with work days within the parameters established for one or more other classes of 

employees in the SEC (excluding aggregate work day requirements).
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Table of Contents 

Evaluation Report Summary: SEC-00093, Santa Susana Field Laboratory-Area IV............................. 3

1.0 Purpose and Scope.......................................................................................................................... 9

2.0 Introduction .................................................................................................................................... 9

3.0 Petitioner-Requested Class/Basis & NIOSH-Proposed Class/Basis ............................................ 10

4.0 Data Sources Reviewed by NIOSH.............................................................................................. 11 

4.1 Site Profile and Technical Basis Documents (TBDs) ......................................................... 11 

4.2 ORAU Technical Information Bulletins (OTIBs) and Procedures ..................................... 12 

4.3 Facility Employees and Experts .......................................................................................... 13 

4.4 Previous Dose Reconstructions ........................................................................................... 13 

4.5 NIOSH Site Research Database .......................................................................................... 14 

4.6 Documentation and/or Affidavits Provided by Petitioners ................................................. 14

5.0 Radiological Operations Relevant to the Class Evaluated by NIOSH ......................................... 17 

5.1 SSFL-Area IV Plant and Process Descriptions ................................................................... 18 

5.2 SSFL-Area IV Functional Areas ......................................................................................... 18 

5.2.1 SSFL-Area IV Nuclear Reactor Development and Testing .................................... 19 

5.2.1.1 Homogeneous Water Boiler Reactors ...................................................... 19 

5.2.1.2 Sodium-Cooled Graphite-Moderated Reactor.......................................... 19 

5.2.1.3 Uranium-Zirconium Hydride Reactors..................................................... 20 

5.2.1.4 Critical Test Facilities............................................................................... 21 

5.2.2 SSFL-Area IV Nuclear Support Operations............................................................ 22 

5.2.2.1 Reactor Fuel Manufacturing..................................................................... 23 

5.2.2.2 Disassembly and Examination of Reactors and Used Reactor Fuel 

Assemblies................................................................................................ 23 

5.2.2.3 Fabrication, Use, and Storage of Radioactive Sources............................. 23 

5.2.2.4 Preparation of Radioactive Material for Disposal .................................... 24 

5.2.2.5 Research on Reprocessing Used Reactor Fuel ......................................... 24 

5.2.2.6 Operation of Particle Accelerators ........................................................... 24 

5.2.2.7 Research Using Radioisotopes ................................................................. 25 

5.2.2.8 Miscellaneous Operations ........................................................................ 26 

5.2.3 Summary of Key SSFL-Area IV Facilities ............................................................. 26 

5.3 Radiological Exposure Sources from SSFL-Area IV Operations ....................................... 27 

5.3.1 Alpha ....................................................................................................................... 30 

5.3.2 Beta/Photon ............................................................................................................. 30 

5.3.3 Neutron.................................................................................................................... 31 

5.3.4 Incidents and Fires................................................................................................... 32

6.0 Summary of Available Monitoring Data for the Class Evaluated by NIOSH ............................. 33 

6.1 SSFL-Area IV Internal Monitoring Data ............................................................................ 33 

6.1.1 Internal Urinalysis Data........................................................................................... 33 

6.1.2 Lung, Whole-Body, and Other Types of Internal or Bioassay Data ....................... 35 

6.2 SSFL-Area IV External Monitoring Data ........................................................................... 35 
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6.3 SSFL-Area IV Air Sampling Data ...................................................................................... 36

7.0 Feasibility of Dose Reconstruction for the Class Evaluated by NIOSH ...................................... 37 

7.1 Pedigree of SSFL-Area IV Data.......................................................................................... 38 

7.1.1 Internal Monitoring Data Review............................................................................ 39 

7.1.2 External Monitoring Data Review........................................................................... 39 

7.2 Internal Radiation Doses at SSFL-Area IV......................................................................... 40 

7.2.1 Process-Related Internal Doses at SSFL-Area IV................................................... 41 

7.2.1.1 Urinalysis Information and Available Data.............................................. 41 

7.2.2 Ambient Environmental Internal Radiation Doses at SSFL-Area IV ..................... 42 

7.2.3 Internal Dose Reconstruction .................................................................................. 42 

7.2.4 Internal Dose Reconstruction Feasibility Conclusion ............................................. 44 

7.3 External Radiation Doses at SSFL-Area IV........................................................................ 44 

7.3.1 Process-Related External Radiation Doses at SSFL-Area IV ................................. 45 

7.3.1.1 Beta and Photon Radiation Dosimetry and Source Term......................... 45 

7.3.1.2 Neutron Radiation Dosimetry and Source Term...................................... 46 

7.3.1.3 Application of Coworker Data for External Dose Reconstruction........... 47 

7.3.2 Ambient Environmental External Radiation Doses at SSFL .................................. 47 

7.3.3 SSFL-Area IV Occupational X-Ray Examinations................................................. 48 

7.3.4 External Dose Reconstruction ................................................................................. 48 

7.3.4.1 Photon Dose.............................................................................................. 49 

7.3.4.2 Electron Dose ........................................................................................... 49 

7.3.4.3 Neutron Dose............................................................................................ 49 

7.3.4.4 Medical X-ray........................................................................................... 49 

7.3.5 External Dose Reconstruction Feasibility Conclusion............................................ 49 

7.4 Evaluation of Petition Basis for SEC-00093....................................................................... 49 

7.4.1 Evaluation of Specific Petitioner Statements in SEC-00093 .................................. 50 

7.4.1.1 SRE Incident and Release of Core Gases................................................. 50 

7.4.1.2 Radiation Badges...................................................................................... 50 

7.4.1.3 Tritium Plumes ......................................................................................... 51 

7.4.1.4 Uranium Fires........................................................................................... 51 

7.4.1.5 Air Monitoring.......................................................................................... 52 

7.4.1.6 The Sodium Burn Pit................................................................................ 52 

7.5 Other Issues Relevant to the Petition Identified During the Evaluation ............................. 53 

7.6 Summary of Feasibility Findings for Petition SEC-00093.................................................. 54

8.0 Evaluation of Health Endangerment for Petition SEC-00093...................................................... 55

9.0 NIOSH-Proposed Class for Petition SEC-00093 ......................................................................... 55

10.0 References .................................................................................................................................... 58 

Tables 

Table 4-1: No. of SSFL-Area IV Claims Submitted Under the Dose Reconstruction Rule ................. 14

Table 5-1: Key SSFL-Area IV Facilities, Operations, and Dates of Operation .................................... 26 
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Table 5-2: Area Information and Parameters of SSFL-Area IV Nuclear Operations ........................... 28 

Table 5-3: Beta and Photon Energies and Percentages at Reactors and Critical Facilities................... 31 

Table 5-4: Facilities and Neutron Energies........................................................................................... 32

Table 6-1: Internal Dose Control Program (in vitro)............................................................................. 33 

Table 6-2: Detection Limits 1958–1966a ............................................................................................. 34 

Table 6-3: Estimated Exchange Frequencies, MDLs, and Annual Missed Doses ................................ 36

Table 7-1: Summary of Feasibility Findings for SEC-00093 ............................................................... 54 
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SEC Petition Evaluation Report for SEC-00093

ATTRIBUTION AND ANNOTATION: This is a single-author document. All conclusions drawn from 

the data presented in this evaluation were made by the ORAU Team Lead Technical Evaluator: Ray 

Clark, Oak Ridge Associated Universities. These conclusions were peer-reviewed by the individuals 

listed on the cover page. The rationales for all conclusions in this document are explained in the 

associated text.

1.0 Purpose and Scope 

This report evaluates the feasibility of reconstructing doses for all employees who worked in any area 

of Santa Susana Field Laboratory-Area IV during the time period from January 1, 1955 to December 

31, 1965. It provides information and analyses germane to considering a petition for adding a class of 

employees to the congressionally-created SEC. 

This report does not make any determinations concerning the feasibility of dose reconstruction that 

necessarily apply to any individual energy employee who might require a dose reconstruction from 

NIOSH. This report also does not contain the final determination as to whether the proposed class 

will be added to the SEC (see Section 2.0). 

This evaluation was conducted in accordance with the requirements of EEOICPA, 42 C.F.R. pt. 83, 

and the guidance contained in the Office of Compensation Analysis and Support’s (OCAS) Internal 

Procedures for the Evaluation of Special Exposure Cohort Petitions, OCAS-PR-004. 

2.0 Introduction 

Both EEOICPA and 42 C.F.R. pt. 83 require NIOSH to evaluate qualified petitions requesting that the 

Department of Health and Human Services (HHS) add a class of employees to the SEC. The 

evaluation is intended to provide a fair, science-based determination of whether it is feasible to 

estimate with sufficient accuracy the radiation doses of the class of employees through NIOSH dose 

reconstructions.1

 

42 C.F.R. § 83.13(c)(1) states: Radiation doses can be estimated with sufficient accuracy if NIOSH 

has established that it has access to sufficient information to estimate the maximum radiation dose, 

for every type of cancer for which radiation doses are reconstructed, that could have been incurred in 

plausible circumstances by any member of the class, or if NIOSH has established that it has access to 

sufficient information to estimate the radiation doses of members of the class more precisely than an 

estimate of the maximum radiation dose.

 

Under 42 C.F.R. § 83.13(c)(3), if it is not feasible to estimate with sufficient accuracy radiation doses 

for members of the class, then NIOSH must determine that there is a reasonable likelihood that such 

radiation doses may have endangered the health of members of the class The regulation requires 

NIOSH to assume that any duration of unprotected exposure may have endangered the health of 

 

1

 NIOSH dose reconstructions under EEOICPA are performed using the methods promulgated under 42 C.F.R. pt. 82 and 

the detailed implementation guidelines available at http://www.cdc.gov/niosh/ocas. 
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members of a class when it has been established that the class may have been exposed to radiation 

during a discrete incident likely to have involved levels of exposure similarly high to those occurring 

during nuclear criticality incidents. If the occurrence of such an exceptionally high-level exposure has 

not been established, then NIOSH is required to specify that health was endangered for those workers 

who were employed for at least 250 aggregated work days within the parameters established for the 

class or in combination with work days within the parameters established for other SEC classes 

(excluding aggregate work day requirements). 

NIOSH is required to document its evaluation in a report, and to do so, relies upon both its own dose 

reconstruction expertise as well as technical support from its contractor, Oak Ridge Associated 

Universities (ORAU). Once completed, NIOSH provides the report to both the petitioner(s) and to the 

Advisory Board on Radiation and Worker Health (Board). The Board will consider the NIOSH 

evaluation report, together with the petition, petitioner(s) comments, and other information the Board 

considers appropriate, in order to make recommendations to the Secretary of HHS on whether or not 

to add one or more classes of employees to the SEC. Once NIOSH has received and considered the 

advice of the Board, the Director of NIOSH will propose a decision on behalf of HHS. The Secretary 

of HHS will make the final decision, taking into account the NIOSH evaluation, the advice of the 

Board, and the proposed decision issued by NIOSH. As part of this decision process, petitioners may 

seek a review of certain types of final decisions issued by the Secretary of HHS.2

 

3.0 Petitioner-Requested Class/Basis & NIOSH-Proposed Class/Basis 

Petition SEC-00093, qualified on October 19, 2007, requested that NIOSH consider the following 

class for addition to the SEC: All employees who worked in all areas of Santa Susana Field 

Laboratory-Area IV during the time period from 1955 to the present (which incorporated the post1987 remediation period). 

The petitioner provided information and affidavit statements in support of the petitioner’s belief that 

accurate dose reconstruction over time is impossible for the Santa Susana Field Laboratory (SSFL)-

Area IV workers in question. NIOSH deemed the following information and affidavit statements 

sufficient to qualify SEC-00093 for evaluation: 

As identified in Item F.1 of the SEC-00093 Form B and discussed in the July 30, 2007 call, 

the petitioner discussed the sodium burn pit, the lack of internal monitoring data, and 

indicated that no records were kept. The petitioner also discussed a Tiger Team report, 

indicating that the report detailed “inadequate air monitoring” and that “no internal 

monitoring was done.” 

 

The information and statements regarding the lack of pre-1959 internal monitoring data qualified the 

petition for further consideration by NIOSH, the Board, and HHS. NIOSH determined that the time 

period from 1955 to 1965 should qualify for the purposes of the SEC petition evaluation, which 

includes the employment period in the original SEC-00093 petition. The details of the petition basis 

are addressed in Section 7.4. 

 

2

 See 42 C.F.R. pt. 83 for a full description of the procedures summarized here. Additional internal procedures are 

available at http://www.cdc.gov/niosh/ocas. 
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Based on its research, NIOSH reduced the petitioner-requested class to define a single class of 

employees for which NIOSH cannot estimate radiation doses with sufficient accuracy. The revised 

NIOSH-proposed class includes all employees who worked in all areas of Area IV of the Santa Susana 

Field Laboratory from January 1, 1955 through December 31, 1958. The petitioner-requested class 

was reduced in scope as a result of the petition qualification review discussed above and the feasibility 

evaluation documented in Section 7.0. With the exception of the class above, NIOSH has obtained 

sufficient information to support bounding doses for the entire evaluation period. 

NIOSH cannot estimate internal exposures with sufficient accuracy during the period from 1955 

through 1958 (which is a period with limited internal monitoring data). This includes the time from 

the beginning of Area IV radiological activities in 1955, to late 1958 (selected as December 31, 1958 

for SEC evaluation purposes). This timeframe corresponds with the date after which an established 

bioassay program existed at SSFL and after which sufficient internal monitoring has been identified. 

In its original evaluation report (Rev. 0) for SEC00093, the NIOSH proposed class definition to 

included all employees of the Department of Energy (DOE), its predecessor agencies, and DOE 

contractors and subcontractors who were monitored while working in any area of Area IV of the Santa 

Susana Field Laboratory from January 1, 1955 through December 31, 1958. 

 

The class definition proposed by NIOSH in revision 0 was based on the infeasibility of reconstructing 

internal doses for the period from January 1, 1955 to December 31, 1958. Information available at the 

time Revision 0 of the report was completed indicated that all workers with the potential for internal 

exposures monitored with a radiation dosimeter badge. The change in proposed class definition was 

triggered by additional information collected by NIOSH indicating that there was a possibility that not 

all workers who entered controlled areas were necessarily monitored for external radiation exposure. 

Because of this, NIOSH can not definitively rule out the possibility that some unmonitored workers 

could have incurred internal exposures. 

4.0 Data Sources Reviewed by NIOSH 

NIOSH identified and reviewed numerous data sources to obtain information relevant to determining 

the feasibility of dose reconstruction for the class of employees proposed for this petition. This 

included determining the availability of information on personal monitoring, area monitoring, 

industrial processes, and radiation source materials. The following subsections summarize the data 

sources identified and reviewed by NIOSH. 

4.1 Site Profile and Technical Basis Documents (TBDs) 

A Site Profile provides specific information concerning the documentation of historical practices at 

the specified site. Dose reconstructors can use the Site Profile to evaluate internal and external 

dosimetry data for monitored and unmonitored workers, and to supplement, or substitute for, 

individual monitoring data. A Site Profile consists of an Introduction and five Technical Basis 

Documents (TBDs) that provide process history information, information on personal and area 

monitoring, radiation source descriptions, and references to primary documents relevant to the 

radiological operations at the site. The Site Profile for a small site may consist of a single document. 

As part of NIOSH’s evaluation detailed herein, it examined the following TBDs for insights into 

SSFL-Area IV operations or related topics/operations at other sites: 
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• Atomics International – Introduction, ORAUT-TKBS-0038-1; Rev. 01; August 30, 2006; SRDB 

Ref ID: 30080 

• Energy Technology Engineering Center – Site Description, ORAUT-TKBS-0038-2; Rev. 00; 

February 2, 2006; SRDB Ref ID: 22140 

• Area IV of the Santa Susana Field Laboratory, the Canoga Avenue Facility (Vanowen Building), 

the Downey Facility, and the De Soto Avenue Facility (sometimes referred to as Energy 

Technology Engineering Center)– Occupational Medical Dose, ORAUT-TKBS-0038-3; Rev. 02; 

October 31, 2008; SRDB Ref ID: 53184 

• Area IV of the Santa Susana Field Laboratory, the Canoga Avenue Facility (Vanowen Building), 

the Downey Facility, and the De Soto Avenue Facility (sometimes referred to as Energy 

Technology Engineering Center [ETEC] or Atomics International) – Occupational Environmental 

Dose, ORAUT-TKBS-0038-4; Rev. 01; March 8, 2007; SRDB Ref ID: 30622 

• Energy Technology Engineering Center – Occupational Internal Dose, ORAUT-TKBS-0038-5; 

Rev. 00; February 22, 2006; SRDB Ref ID: 22287 

• Atomics International – Occupational External Dose, ORAUT-TKBS-0038-6; Rev. 01; November 

16, 2006; SRDB Ref ID: 30082 

4.2 ORAU Technical Information Bulletins (OTIBs) and Procedures 

An ORAU Technical Information Bulletin (OTIB) is a general working document that provides 

guidance for preparing dose reconstructions at particular sites or categories of sites. An ORAU 

Procedure provides specific requirements and guidance regarding EEOICPA project-level activities, 

including preparation of dose reconstructions at particular sites or categories of sites. NIOSH 

reviewed the following OTIBs and procedures as part of its evaluation: 

• OTIB: Estimating the Maximum Plausible Dose to Workers at Atomic Weapons Employer 

Facilities; ORAUT-OTIB-0004; Rev. 03 PC-2; December 6, 2006; SRDB Ref ID: 29949

• OTIB: Dose Reconstruction from Occupationally Related Diagnostic X-Ray Procedures; ORAUTOTIB-0006; Rev. 03; August 2, 2005; SRDB Ref ID: 19422

• OTIB: A Standard Complex-Wide Methodology for Overestimating External Doses Measured with 

Film Badge Dosimeters; ORAUT-OTIB-0010; Rev. 01; June 5, 2006; SRDB Ref ID: 29953

• OTIB: Interpretation of Dosimetry Data for Assignment of Shallow Dose; ORAUT-OTIB-0017; 

Rev. 00; January 19, 2005; SRDB Ref ID: 19435

• OTIB: Internal Dose Overestimates for Facilities with Air Sampling Programs; ORAUT-OTIB0018; Rev. 01; August 9, 2005; SRDB Ref ID: 19436
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• OTIB: Use of Coworker Dosimetry Data for External Dose Assignment; ORAUT-OTIB-0020; 

Rev. 0; December 29, 2004; SRDB Ref ID: 19441

• OTIB: Application of Internal Doses Based on Claimant-Favorable Assumptions for Processing as 

Best Estimates; Rev. 00; ORAUT-OTIB-0033; April 20, 2005; SRDB Ref ID: 19457

• OTIB: Fission and Activation Product Assignment for Internal Dose-Related Gross Beta and Gross 

Gamma Analyses; ORAUT-OTIB-0054; Rev. 00; May 11, 2007; SRDB Ref ID: 31780

• OTIB: Technical Basis for Conversion from NCRP Report 38 Neutron Quality Factors to ICRP 

Publication 60 Radiation Weighting Factors for Respective IREP Input Neutron Energy Ranges; 

ORAUT-OTIB-0055; Rev. 00; June 5, 2006; SRDB Ref ID: 29980

• Procedure: Occupational Onsite Ambient Dose Reconstruction for DOE Sites, ORAUT-PROC0060; Rev. 01; June 28, 2006; SRDB Ref ID: 29986

• Procedure: Occupational X-Ray Dose Reconstruction for DOE Sites, ORAUT-PROC-0061; Rev. 

01; July 21, 2006; SRDB Ref ID: 29987

4.3 Facility Employees and Experts 

To obtain additional information, NIOSH interviewed five former SSFL employees. The interviews 

were conducted by two ORAU team members and each interview lasted approximately one hour. 

• Personal Communication, 2007, Personal Communication with Fireman; Telephone Interview by 

ORAU Team; November 15, 2007 9:00 AM PST; SRDB Ref ID: 37532

• Personal Communication, 2007, Personal Communication with Health Physicist; Telephone 

Interview by ORAU Team; November 15, 2007 10:30 AM PST; SRDB Ref ID: 37536

• Personal Communication, 2007, Personal Communication with Health Physicist; Telephone 

Interview by ORAU Team; November 20, 2007 11:00 AM PST; SRDB Ref ID: 37534 

• Personal Communication, 2007, Personal Communication with HP-Radiation Engineer; 

Telephone Interview by ORAU Team; November 29, 2007 1:00 PM PST; SRDB Ref ID: 37537 

• Personal Communication, 2007, Personal Communication with Health and Safety Manager; 

Telephone Interview by ORAU Team; November 30, 2007 9:00AM PST; SRDB Ref ID: 37538 

4.4 Previous Dose Reconstructions 

NIOSH reviewed its NIOSH OCAS Claims Tracking System (NOCTS) to locate EEOICPA-related 

dose reconstructions that might provide information relevant to the petition evaluation. Table 4-1 

summarizes the results of this review. (NOCTS data available as of April 21, 2009) 
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Table 4-1: No. of SSFL-Area IV Claims Submitted Under the Dose Reconstruction Rule 

Description Totals 

Total number of claims submitted for dose reconstruction 247 

Total number of claims submitted for energy employees who meet the evaluated class definition for 

the time period evaluated in this report (January 1, 1955 through December 31, 1965) 171 

Number of dose reconstructions completed for energy employees who meet the evaluated class 

definition criteria 118 

Number of claims for which internal dosimetry records were obtained for the identified years in the 

evaluated class definition 28* 

Number of claims for which external dosimetry records were obtained for the identified years in the 

evaluated class definition 63 

Notes: 

* One claim with pre-1959 internal monitoring data (available for September 1958); no internal data available 

prior to the initiation of an internal monitoring program for Area IV of the SSFL (program initiated in August 

1958). 

NIOSH reviewed each claim to determine whether internal and/or external personal monitoring 

records could be obtained for the employee. Of the 171 claims that are within the period of this 

report, 166 have received responses to requests for site data. Of the 118 claims where the associated 

individual dose reconstructions have been completed, all 118 cases have been submitted to DOL. 

4.5 NIOSH Site Research Database 

NIOSH also examined its Site Research Database to locate documents supporting the evaluation of 

the proposed class. Six hundred eighty-eight documents in this database were identified as pertaining 

to Area IV of the Santa Susana Field Laboratory. These documents were evaluated for their relevance 

to this petition. The documents include historical background on internal and external monitoring, 

program descriptions, process materials, process descriptions, dust sampling results, air monitoring 

data, urinalysis data, as well as information about the radiological control program and medical 

monitoring. 

4.6 Documentation and/or Affidavits Provided by Petitioners 

In qualifying and evaluating the petition, NIOSH reviewed the following documents submitted by the 

petitioners: 

• Petition Form B; June 22, 2007; OSA Ref ID: 103252

• Letter to Substantiate SEC Petition Application for the Santa Susana Field Laboratory; [Name 

One Redacted]; June 30, 2007; OSA Ref ID: 103255

• Under Pressure, DOE Halts Field Lab Cleanup for 45 Days, newspaper article submitted with 

petition; Kerry Cavanaugh; May 26, 2007; OSA Ref ID: 103253

• Order Granting Plaintiffs’ Motion For Summary Judgments, Natural Resources Defense Council, 

Inc., Committee to Bridge the Gap, and City of Los Angeles (Plaintiffs) versus Department of 

Energy, Spencer Abraham, Secretary, Department of Energy, and Camile Yuan-Soo Hoo, 
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Manager, National Nuclear Security Administration, Oakland Operations (Defendants); United 

States District Court; May 2, 2007; OSA Ref ID: 103253

• Declaration of Arjun Makhijani, expert testimony in legal case; February 12, 2004; OSA Ref ID: 

103253

• Living on Earth, email containing interview that aired the week of January 20, 2006— containing 

comments by Arjun Makhijani; March 6, 2006; OSA Ref ID: 103253

• Declaration of Bernd Franke, expert testimony in legal case; February 12, 2004; OSA Ref ID: 

103253

• Curriculum Vitae for Bernd Franke; OSA Ref ID: 103253

• Santa Susana Field Lab (SSFL) Closure Plan, letter from Julia Brownley; Julia Brownley, 

Assembly Member of the 41st District; May 4, 2007; OSA Ref ID: 103253

• Statement of the Honorable John N. Hostettler, Chairman of the Subcommittee on Immigration, 

Border Security and Claims for the December 5, 2006 Oversight Hearing on the Energy 

Employees Occupational Illness Compensation Program Act—Are we Fulfilling the Promise We 

Made to These Veterans of the Cold War when We Created the Program?; OSA Ref ID: 103253

• Monthly Report—March, 1960; Chemical Disposal—Activities at Pits, Inspections, Burning, 

Convoying, Etc.; Inter-office letter from E. G. Addeo to R. J. Madden; E. G. Addeo; April 4, 

1960; OSA Ref ID: 103253

• Various Pages from Journal, pictures and pages from article that was published in Ecology and 

Environment, Inc.; author not identified; July 19, 1989; OSA Ref ID: 103253

• EPA Concludes Superfund Evaluation of ETEC Area IV, United States Environmental Protection 

Agency; December 2003; OSA Ref ID: 103253

• Energy Employees Occupational Illness Compensation Program Act of 2000, letter from [Name 

One Redacted] to Congressman Elton Gallegly; [Name One Redacted]; March 1, 2006; OSA Ref 

ID: 103253

• H. R. 2242, copy of H. R. 2243; 110th Congress, 1st Session; May 9, 2007; OSA Ref ID: 103253

• The U.S. EPA Announces Results of Rocketdyne’s Off-Site Sampling Program for the Santa 

Susana Field Laboratory, EPA update; U. S. Environmental Protection Agency; July 1999; OSA 

Ref ID: 103253

• Dose Construction [sic] for [Name Two Redacted], letter from [Name One Redacted] to Mr. Larry 

Elliott; [Name One Redacted]; June 4, 2006; OSA Ref ID: 103253

• Commendation Regarding Uranium Fire, Canoga Facility, May 17, 1967, J. E. Stewart; May 19, 

1967; OSA Ref ID: 103253
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• Employment History/Monitoring Record; Santa Susana Field Laboratory; October 28, 2005; OSA 

Ref ID: 103253

• Statement from [Name Three Redacted], regarding work experiences at Rocketdyne; date not 

identified; OSA Ref ID: 103253

• Site Visit to Santa Susana Field Laboratory Operated by Rockwell/Rocketdyne, site visit memo; 

Gregg D. Dempsey; July 28, 1989; OSA Ref ID: 103253

• Letter Regarding Request to Evaluate the Need for an Epidemiological Study of Workers and 

Community Residents Potentially Impacted by the Operations at Rockwell International 

Corporation’s Santa Susana Field Laboratories, letter to Dr. Robert Goldsmith from the 

Department of Health Services; Department of Health Services of California; October 5, 1990; 

OSA Ref ID: 103253

• Statement of the Potential Health Risks Identified in the Santa Susana Field Laboratory Exposure 

Assessment Study; Charles M. Haskell; February 4, 1991; OSA Ref ID: 103253

• High-Level Radioactivity Found in Groundwater at Rocketdyne Lab; Kerry Cavanaugh; May 19, 

2004; OSA Ref ID: 103253

• Higher Tritium Levels Found in Water at Field Lab; Kerry Cavanaugh; 2005; OSA Ref ID: 

103253

• Studies of Atomic Bomb Survivors—Understanding Radiation Effects, Editorial in the Journal of 

the American Medical Association, John D. Boice; August 1, 1990; OSA Ref ID: 103253

• Excerpts from Unknown Report; unknown author; unknown date; OSA Ref ID: 103253, pages 151 

and 154

• Project History Excerpts from Unknown Report; unknown author; unknown date; OSA Ref ID: 

103253, pages 155 and 156

• Dispute Surfaces on Rocketdyne Deaths Study, newspaper article; Mack Reed; July 11, 1995; OSA 

Ref ID: 103253

• Rocketdyne Worker Health Study Summary; California Department of Health Services, 

Occupational Health Branch; date not identified; OSA Ref ID: 103253

• Lab Linked to Cancer Risk, article in the Los Angeles Times; Mack Reed and Thomas H. Maugh; 

September 12, 1997; OSA Ref ID: 103253

• Incident Report Regarding Oxidation Reaction of Uranium; D. E. Owens; October 27, 1964; OSA 

Ref ID: 103700
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• Radiation Safety Progress for November 1966, regarding a sodium reactor experiment; author 

unknown; December 9, 1966; OSA Ref ID: 103701

• Letter Regarding SSFL-Area IV Dose Reconstruction Case, includes dose reconstruction with 

attached DR Report and handwritten statements/notes; [Name One Redacted]; September 6, 2007; 

OSA Ref ID: 103702

• Interim Final RCRA Facility Assessment Report for Rockwell International Corporation, TZ4-

R09015-RN-M07933; Science Applications International Corporation, Technology Services 

Company; July 10, 1991; OSA Ref ID: 103715

• Email from Petitioner, sent to OCAS requesting document be added to petition; [Name One 

Redacted]; August 27, 2007; OSA Ref ID: 103666

• Tiger Team Assessment—Energy Technology Engineering Center, DOE/EH—0175; U.S. 

Department of Energy; April 1991; OSA Ref ID: 103714

• Detailed Outline of Materials in Support of SEC-00093, Email to OCAS; [Name One Redacted]; 

September 20, 2007; OSA Ref ID: 103808

• The Rocket’s Red Glare, a short DVD from Rocketdynewatch.org

• A Briefing for Elected Officials on the Health Risks to the Public from SSFL, DVD submitted by 

petitioner, dated April 5, 2007

5.0 Radiological Operations Relevant to the Class Evaluated by 

NIOSH 

Unless otherwise indicated, information for Section 5.0 and its subsections was obtained from

Historical Site Assessment of Area IV Santa Susana Field Laboratory Ventura County, California, 

Volume 1-Methodology, Nuclear Operations at Rockwell’s Santa Susana Field Laboratory—A 

Factual Perspective, or Energy Technology Engineering Center – Site Description (ORAUT-TKBS0038-2). 

The following subsections summarize both radiological operations at SSFL-Area IV and the 

information available to NIOSH to characterize particular processes and radioactive source materials. 

From available sources NIOSH has gathered process and source descriptions, information regarding 

the identity and quantities of each radionuclide of concern, and information describing both processes 

through which radiation exposures may have occurred and the physical environment in which they 

may have occurred. The information included within this evaluation report is intended only to be a 

summary of the available information. 

For the purpose of providing a complete understanding of the history of SSFL-Area IV, Section 5.0 

includes discussions about operations that occurred outside of the period being evaluated in this 

report. 
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5.1 SSFL-Area IV Plant and Process Descriptions 

SSFL consists of a total of 2,850 acres and is located in the Simi Hills of Ventura County, 

approximately 30 miles northwest of downtown Los Angeles, California. Based on ownership and 

operations, SSFL is divided into four administrative and operational portions—Area I, Area II, Area 

III, and Area IV. DOE operations are conducted in Rockwell International-owned and DOE-owned 

facilities on a 290-acre westernmost administrative and operational portion designated as Area IV. 

Following World War II, the potential of atomic energy captured the interest of the United States 

Government and many companies. This interest resulted in the need for Nuclear Research & 

Development (R&D) facilities. SSFL was initially established in 1947 by North American Aviation 

(NAA) to meet the requirements for a field test laboratory to static-fire large rocket engines; however, 

it also met NAA’s need for a Nuclear Research facility. As a result, Area IV was established in 1953 

at SSFL as a Nuclear Research and Development facility. Since then, SSFL-Area IV has housed both 

nuclear development and rocket development groups, albeit in distinct and separate locations. Atomic 

Energy Research Development (AERD) also conducted operations in SSFL-Area IV. In December 

1955, the nuclear development and rocket development groups were transformed into separate 

divisions: Atomics International (AI) and Rocketdyne.

Two distinct AI groups were housed in Area IV and supported by DOE. One focused on development 

of civilian nuclear power, and the other was a center of excellence for research and testing of nonnuclear components related to liquid metals. These two groups were referred to as AI and Liquid 

Metal Engineering Center (LMEC), respectively. Nuclear R&D activities in Area IV increased 

rapidly from 1953 into the late 1960s, and then began to decline. AI was eventually merged into 

Rocketdyne in 1984 as a result of this decline. 

The LMEC was created in 1966 as a government-owned and contractor-operated organization; its 

purpose was to provide development and non-nuclear testing of Liquid Metal Reactor (LMR) 

components and to establish the Liquid Metal Information Center (LMIC) for the AEC’s Liquid Metal 

Fast-Breeder Reactor (LMFBR) program. The LMEC was renamed Energy Technology Engineering 

Center (ETEC) in 1978 to reflect DOE’s desire to broaden its mission beyond the LMFBR program. 

Several corporate mergers and organizational changes occurred over the years. In 1967, NAA merged 

with Rockwell Standard to become North American Rockwell. In 1973, the corporate name changed 

to Rockwell International (RI), with AI and Rocketdyne continuing to exist as independent divisions 

until 1984, when AI was absorbed by the Rocketdyne division. The Boeing Company purchased RI 

in 1996, and Rocketdyne is now a division of Boeing. 

The DOE Office of Worker Advocacy defines the start of covered operations as 1955 for Area IV of 

the Santa Susana Field Laboratory; it identifies the Downey Facility, DeSoto Avenue Facility, and 

Canoga Avenue Facility (all located outside the boundary of the SSFL site) as separate covered 

facilities under EEOICPA. Therefore, this report only includes a review and evaluation of the covered 

periods and locations as defined by the DOE Office of Worker Advocacy for Area IV of the SSFL. 

5.2 SSFL-Area IV Functional Areas 

SSFL-Area IV operations included the following functional areas: 
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• Nuclear Reactor Development and Testing 

• Nuclear Support Operations 

Additional information regarding the functional areas of SSFL-Area IV can also be found in ORAUTTKBS-0038-2. 

5.2.1 SSFL-Area IV Nuclear Reactor Development and Testing 

Between 1954 and 1980, several nuclear reactors were built, tested, and operated in Area IV. These 

included both nuclear reactors and critical test assemblies. Nuclear reactor programs focused on the 

development and operation of homogeneous water boiler-type reactors, sodium-cooled graphite 

moderated reactors, and uranium-zirconium hydride reactors. 

Operation of a nuclear reactor creates three sources of radioactivity: fission, transuranic, and 

activation products. When part of the fissionable material in the fuel element is expended, or when a 

reactor is decommissioned, the fuel elements are removed. These spent fuel elements contain the 

fission products and transuranic materials generated by operation of the reactor, and the activation 

products in the cladding. The amount of radioactivity generated by a nuclear reactor depends in part 

on the amount of heat it generates. The heat generated is measured in thermal watts (Wt). The 

reactors at SSFL all operated at low-power levels. Six had power levels of less than 100 kWt, three 

had power levels of 600 to 1,000 kWt, and one was a 20-MWt test reactor. 

A total of 135 million Ci of radioactivity was generated in the fuel from about 7,200 MWd of total 

reactor operation. Almost 90% of the total SSFL radioactivity was generated by the 20-MWt Sodium 

Reactor Experiment (SRE). 

5.2.1.1 Homogeneous Water Boiler Reactors

The water boiler reactors were operated in Buildings 4073 and 4093. The water boiler reactors used a 

93% enriched uranyl sulfate solution held in a critical configuration in a spherical vessel. Rather than 

actually boil, the neutron and gamma flux caused radiolytic decomposition of water into hydrogen and 

oxygen in the form of tiny bubbles, which gave the impression of boiling. Area IV contained two 

water boiler reactors: the Kinetics Experiment Water Boiler (KEWB) operated from 1956 until 1966 

and the Water Boiler Neutron Source (WBNS), later called the L-85, operated from 1956 through 

1980. 

5.2.1.2 Sodium-Cooled Graphite-Moderated Reactor

The Sodium-Cooled Graphite-Moderated Reactor in Area IV of the SSFL was designated as the 

Sodium Reactor Experiment (SRE). The SRE included 12 structures on the SSFL site including the 

reactor building, office buildings, and support structures. Eight of these structures were directly 

involved in operations with radioactive materials: 

• Reactor Building (Building 4143) 

• Component Storage Building (Building 4041) 
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• Temporary Hot Waste Storage Building (Building 4686) 

• Site Service Building (Building 4163) 

• Cold Trap Vault (Building 4695) 

• Liquid Radioactive Waste Vault (Building 4653) 

• Interim Radioactive Waste Storage Area (Area 4654) 

• Intermediate Contaminated Storage Area (Area 4689) 

The SRE started in Area IV, Building 4143 in 1954 as part of an AEC program to develop a graphitemoderated sodium-cooled reactor for power application and to demonstrate the feasibility of such 

reactors as an energy source for power stations. Construction of the SRE reactor was completed in 

February 1957, with the reactor being brought to full power in May 1958; it operated until February 

1964, when DOE terminated the program (having generated 37,000 MWhr of electrical power in more 

than 27,000 operating hours during this time). Decontamination and decommissioning (D&D) of the 

SRE began in 1974 and was completed in 1983. 

5.2.1.3 Uranium-Zirconium Hydride Reactors

The Systems for Nuclear Auxiliary Power (SNAP) program operated from 1956 to 1971 to support 

the development and testing of small reactors designed to provide power for research missions in 

space. The SNAP reactors were uranium-zirconium hydride reactors that used fully-enriched (93%) 

uranium dispersed in fuel rods containing zirconium hydride. Seven SNAP reactors were tested and 

operated in Buildings 4010, 4024, 4028, and 4059. The seven SNAP reactors included the following: 

SNAP Experimental Reactor (SER), SNAP 8 Experimental Reactor (S8ER), SNAP 2 Development 

Reactor (S2DR), SNAP 10 Flight Simulation Reactor (S10FS3), Shield Test Reactor (STR), Shield 

Test and Irradiation Reactor (STIR), and SNAP 8 Development Reactor (S8DR). 

SNAP Experimental Reactor (SER)—The SER operated from September 1959 through December 

1960, when the reactor and the associated test equipment were removed, examined, and sent offsite 

for disposal. 

SNAP 8 Experimental Reactor (S8ER)—The S8ER operated from May 1963 to April 1965, when 

the core and associated equipment were removed, examined at the Hot Lab, and sent offsite for 

disposal. Building 4010 D&D was completed in 1978 and the building demolished thereafter. 

SNAP 2 Development Reactor (S2DR)—The S2DR operated from April 1961 through December 

1962. The S2DR was tested at a nominal power level of 65 kWt without any power conversion 

system equipment. The reactor was tested in Vault 1 of the SNAP Environmental Test Facility, 

Building 40243

. 

 

3

 Building 4024 also contained the SNAP 10 Flight System 3 (S10FS3) reactor in a different vault. In addition, the SNAP 

Transient (SNAPTRAN) Reactor criticality tests occurred in this building. 
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SNAP 10 Flight Simulation Reactor (S10FS3)—The S10FS3 was a SNAP 10A reactor that operated 

between January 1965 and March 1966; it was used to test the reliability and performance of the 

reactors in space. Following completion of S10FS3 operations, the reactor was removed from the 

building and sent offsite for disposal. The D&D activities for Building 4024 were completed in 1978. 

However, some activated concrete shielding remains in the vaults and the test vaults remain restricted; 

they have been in surveillance and maintenance mode since 1978. 

Shield Test Reactor (STR)—The STR operated from 1961 to 1964 and was fueled with SNAP 

reactor fuel elements. In 1964, the reactor was renamed the Shield Test and Irradiation Reactor 

(STIR). 

Shield Test and Irradiation Reactor (STIR)—The STIR operated from 1964 to 1972. In 1972, fuel 

from the STIR was removed and disposed of; the facility was secured. D&D ended in March 1976. 

SNAP 8 Development Reactor (S8DR)—Building 4059 was built during 1962 and 1963 for 

development testing of SNAP reactors. At the end of test operations, the reactor core and control 

system were removed and shipped offsite for disposal. At this time, sufficient D&D occurred to make 

a portion of the facility available for other use. Further D&D of the facility (except for the reactor cell 

and vacuum system) occurred between June and September, 1978. Upon completion of D&D in 

1992, the entire facility was released for unrestricted use. 

5.2.1.4 Critical Test Facilities

A criticality test operates at a very low-power level (up to a few hundred watts), and neutron levels are 

correspondingly very low (which results in the generation of correspondingly lower amounts of 

radioactivity and radioactive materials, as compared to production-scale operating reactor units that 

operate at much higher power levels and outputs). Therefore, a large number of criticality tests can 

occur in the same test facility without generating much activation product radioactivity. Almost the 

entire amount of generated radioactivity is contained within the fuel elements of the criticality test. 

When the fuel elements are removed, the radioactivity is also removed. There have been dozens of 

criticality tests at Area IV in seven different facilities (i.e., First and Second SNAP Critical Facility, 

SNAP Flight System Critical Facility, SNAP Transient Test Facility, Organic Moderated Reactor, 

Sodium Graphite Reactor, Advanced Epithermal Thorium Reactor, and the Fast Critical Experiment 

Laboratory). Use of some of these low-power reactors began in 1954 and continued until 1974. 

SNAP Development Test Facilities—Test facilities in Area IV for the SNAP program included 

Buildings 4373, 4012, 4019, and 4024. 

First SNAP Critical Facility, Building 4373—The First SNAP Critical Facility was a Critical Test 

Facility constructed in Building 4373. It was used for five SNAP reactor critical assembly tests 

(SCA-1, S2ERC, SCA-2, SCA¬3, and SCA-4C) between 1957 and 1963. The D&D activities in 

Building 4373 were completed in 1995. Demolition of the building was completed in 2003. 

Second SNAP Critical Facility, Building 4012—Three SNAP reactor critical assemblies (SCA-4A, 

SCA-4B, and SCA-5) were tested between 1961 and 1967. Initial demolition efforts in Building 4012 

were completed in 1986. Final D&D in the remaining portion of the building was performed in 1995. 

Demolition of the building was completed in 2003. 
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SNAP Flight System Critical Facility, Building 4019—The SNAP Flight System Critical Facility was 

built to test-qualify SNAP reactor power systems. Additionally, three reactors (FS-1, FS-4, and FS-5) 

were assembled and tested from 1964 through 1965. D&D activities for the SNAP Flight System 

Critical Facility were completed in 1965. Following additional decontamination in 1998, the building 

was released. 

SNAP Transient Test Facility, Building 4024—The SNAP Transient Test Facility only operated in 

1971. 

Note: Information regarding the SNAP Transient Test Facility is included to provide a complete 

background of the processes at SSFL—Area IV. However, based on the operational timeframe of the 

facility, anything related to the SNAP Transient Test Facility is considered outside the scope of this 

evaluation. 

Civilian Nuclear Power Test Facilities—Critical test facilities supporting the development of 

civilian nuclear power included Buildings 4009 and 4100. 

Organic Moderated Reactor (OMR), Building 4009—The OMR was used from 1958 through 1967. 

D&D activities for the OMR were completed in 1967. When the program ended, all associated 

equipment was removed. Building 4009 is currently being used for non-nuclear R&D. 

Sodium Graphite Reactor (SGR), Building 4009—The SGR operated from 1958 through 1967. D&D 

activities were completed in 1967. When the program ended, all associated equipment was removed. 

Building 4009 is currently being used for non-nuclear R&D. 

Advanced Epithermal Thorium Reactor (AETR), Building 4100—The AETR operated from 1960 

through 1974. D&D activities for the AETR were completed after the program was terminated in 

1974. 

Fast Critical Experiment Laboratory, Building 4100—The Fast Critical Experiment Laboratory was a 

licensed facility used between 1961 and 1974. The laboratory was decontaminated, decommissioned, 

and released for unrestricted use in 1980. The facility license was terminated October 1, 1980. 

5.2.2 SSFL-Area IV Nuclear Support Operations 

Starting in 1956, several operations were conducted in Area IV to support nuclear programs. 

Operations included reactor fuel manufacturing; disassembly and examination of reactors and used 

reactor fuel assemblies; fabrication, use, and storage of radioactive sources; preparation of radioactive 

material for disposal; research on reprocessing used reactor fuel; operation of particle accelerators; 

research using radioisotopes; and other miscellaneous operations, which included the conservation 

yard and sodium disposal facility. With the exception of the Radiation Materials Handling Facility, 

Fuel Storage Facility, and the Radiation Instrument Calibration Laboratory, operations were 

terminated by 1988. Operations were terminated at the Fuel Storage Facility and Radiation 

Instrument Calibration Laboratory in 1993 and 1996, respectively. The RMHF continues to operate in 

support of the DOE’s cleanup efforts. 
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5.2.2.1 Reactor Fuel Manufacturing

As part of the nuclear reactor development work performed for the government, two different reactor 

fuel manufacturing operations occurred at SSFL-Area IV in Buildings 4003 and 4064 during the 

timeframe being evaluated in this report. 

Note: Reactor fuel manufacturing and studies were also performed in Building 4055 (used as a 

Plutonium Fuel Manufacturing Facility) and Building 4005 (used as a Uranium Carbide Fuel 

Manufacturing Pilot Plant); however, radiological activities associated with both of these buildings 

were performed outside of the proposed class timeframe evaluated in this report. Therefore, further 

discussion of these facilities is not included here. 

Engineering Test Building, Building 4003—The Engineering Test Building was built to support the 

SRE through assembly of SRE fuel elements; it operated between 1954 and 1964. D&D was 

completed in 1975. The facility has been released for unrestricted use. 

Fuel Storage Facility, Building 4064—Fuel Storage Facility operations started in 1958. D&D 

activities for Building 4064 were completed in 1997. 

5.2.2.2 Disassembly and Examination of Reactors and Used Reactor Fuel Assemblies

The Hot Lab operated in Building 4020 from 1959 through 1990 and was used to examine fuel and/or 

components from the SRE, SER, S2DR, S8ER, S8DR, and S10FS3 reactors, the OMR and SGR 

Criticality Test Facilities, and the Piqua reactor. When a reactor operation was complete and the 

reactor was no longer needed, the reactor was removed from its operating location, disassembled, its 

fuel removed, and the radioactive structure cut into pieces small enough to be shipped for disposal. 

The Hot Lab was also used to declad fuel from the SRE, EBR-I, EBR-II, Hallam, Fermi, and 

Southwest Experimental Fast Oxide Reactor reactors. Additionally, the Hot Lab was used for work 

on radioactive materials generated outside SSFL; this consisted in large part of used reactor fuel from 

other nuclear reactors. The fuel elements were shipped to the Hot Lab and disassembled or separated 

from their cladding material, and the separated materials were then shipped offsite. Finally, the Hot 

Lab was also used for manufacturing sealed sources, for leak checks on sources, and for cutting and 

machining operations that involved cobalt-60. The license for this facility was changed in 1990 to 

limit operations to D&D work. D&D in Building 4020 was completed in 1999. 

5.2.2.3 Fabrication, Use, and Storage of Radioactive Sources

Building 4029, the Radiation Measurements Facility, housed, stored, and used radioactive sources to 

calibrate radiation detection instruments for the SRE and other reactor tests. Building 4029 operated 

between 1959 and 1974, after which it was used for temporary storage of non-radioactive materials. 

In addition to calibrating nuclear instruments, radiation sources were used for some forms of 

radiography, irradiation testing, and other applications. Sources were manufactured in the Hot Lab 

and used in various facilities both at the SSFL and elsewhere. Approximately 140,000 Ci of 

radioactive material (primarily promethium-147) were fabricated into sources at the Hot Lab. These 

sources were stored in secured locations and used under carefully controlled conditions. 
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5.2.2.4 Preparation of Radioactive Material for Disposal

Radioactive waste from the operation of nuclear reactors, fuel fabrication, reactor and fuel 

examination, etc., was prepared for disposal at the Radioactive Materials Disposal Facility (RMDF) 

with support at the Interim Storage Facility (ISF) in Building 4654. 

Radioactive Materials Disposal Facility (RMDF), Building 4021/4022—The RMDF was built in 

1958 for storing fuel and processing solid and liquid waste for disposal in conjunction with the 

operation of the SRE. The RMDF began operations in 1959 (Sapere Consulting, 2005) and has 

subsequently been used to support all SSFL nuclear operations. The RMDF consisted of the 

following structures and/or areas: 

• Building 4022—Radioactive material storage vault 

• Building 4021—Decontamination and Packaging Facility 

• Building 4075—Low specific activity waste storage 

• Building 4621—Source storage 

• Buildings 4034 & 4044—Offices 

• Building 4665—Non-radioactive material storage 

• Leach Field—Sanitary sewer septic tank 

Interim Storage Facility (ISF), Building 4654—The ISF was built in 1958 and used from 1958 

through 1964 to store SRE fuel elements inside 10 thimbles installed in holes drilled into bedrock 

(ORAUT-TKBS-0038-2). The ISF was also used for storing fuel shipping casks for other reactors 

(OMRE and SNAP). In 1964, the ISF was taken out of service, the thimbles removed, and the area 

decontaminated and released for unrestricted use. 

5.2.2.5 Research on Reprocessing of Used Reactor Fuel

The Hot Cave in Building 4003 (Engineering Test Building) supported testing of nuclear reactor fuel 

reprocessing. From 1957 through 1964, Building 4003 was used to assemble fuel for the SRE. Until 

the termination of the SNAP program in 1973, Building 4003 was used for the analysis of SNAP fuel 

burn-up samples and the evaluation of irradiation experiments (Sapere Consulting, 2005). The used 

fuel assemblies from nuclear reactors contained unused fissionable material, fissionable transuranic 

products (mainly plutonium), and fission products. Rockwell developed a pilot-scale process to make 

a partial separation of used fuel by removing part of the fission products so that the material could be 

used again as reactor fuel. The experiments used up to kilogram quantities of unirradiated uranium 

and thorium, and up to 100-g quantities of highly irradiated materials. 

5.2.2.6 Operation of Particle Accelerators

Besides nuclear fission, radioactivity was also generated by bombarding a target material with atomic 

particles accelerated to high speeds by means of a particle accelerator. Rockwell operated a Van de 
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Graaff generator in Building 4030 from 1960 through 1964, bombarding tritium targets with 

deuterons to produce neutrons (Sapere Consulting, 2005). A second Van de Graaff generator was 

operated at the SRE Facility for neutron activation analyses of materials (removed from the SRE 

Facility before the facility D&D). Accelerators were removed in 1966 and a radiation survey showed 

the facilities to be free of radioactive contamination. A follow-up survey performed in 1988 

confirmed that there was no activation in the building. 

5.2.2.7 Research Using Radioisotopes

Some SSFL-Area IV research required special radioisotopes. Small quantities of specially prepared 

radioisotopes were brought to SSFL-Area IV, used in laboratories under controlled conditions, and 

then either shipped back out or stored for reuse. 

One research program associated with the use of radioisotopes that was planned at SSFL was the 

Transuranic Management by Pyropartitioning-Separation (TRUMP-S) program. The purpose of the 

TRUMP-S program was to develop fundamental thermodynamic and electrochemical data on various 

transuranic materials so that processes could be developed to separate long-lived radioactive isotopes 

from spent nuclear fuel. These long-lived radioactive isotopes could then be destroyed by fissioning 

them in a nuclear reactor or accelerator, thereby eliminating the long-term hazard associated with the 

disposal of spent nuclear fuel. The program used small quantities of transuranic materials (plutonium, 

neptunium, and americium). The TRUMP-S tests were originally planned to be performed in the Hot 

Laboratory at SSFL; however, the test program was transferred to and implemented at the University 

of Missouri. 

Fuel Storage Facility, Building 4064—The Fuel Storage Facility contained the materials for the 

TRUMP-S tests, which were subsequently transferred to the University of Missouri when it was 

determined that the tests were not going to be performed at SSFL. The material that was shipped to 

Missouri included 75 g of depleted uranium, 5 g of plutonium, 4 g of neptunium, and 4 g of 

americium. Building 4064 was constructed in two phases. In 1958, a reinforced concrete structure 

with 11-inch thick walls was constructed. Construction of a second part in 1963 included a new bay 

to the north of the original structure. The facility operated from 1958 until 1993. It was demolished 

in 1997 and the waste was shipped offsite as clean waste (DOE, 2007). 

Corrosion Testing Laboratory, Building 4023— Another radioisotope research program was the 

corrosion test program in the Corrosion Testing Laboratory, which ran from 1962 through 1986. A 

pumped sodium corrosion test loop was built to study the deposition behavior of activation products 

(manganese-54 and cobalt-60) in flowing sodium to develop more effective traps for these isotopes. 

An activated piece of fuel cladding containing these isotopes was used in these tests. The test 

specimens and the test loop were later removed from the building. 

Mechanical Component Development and Counting Building, Building 4363—The Mechanical 

Component Development and Counting Building was transferred from the Rocketdyne Division to 

Area IV of the SSFL between 1956 and 1957 for the purpose of supporting expansion of the nuclear 

operations at SSFL. It was used for sodium systems in support of the SRE from before 1959 until 

1963. It has been primarily used for storage since 1963. 
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5.2.2.8 Miscellaneous Operations

Neither the Conservation Yard nor the Sodium Disposal Facility was intended for use with radioactive 

materials, but both were inadvertently contaminated. 

Conservation Yard—The Conservation Yard was an outdoor area used for storage and salvaging 

used equipment. In 1988, a 20-by-20-foot area of the surface of the Conservation Yard was found to 

be slightly contaminated by radioactivity. A total of 132 cubic feet of contaminated soil and asphalt 

was removed and shipped offsite for disposal. The Conservation Yard was released for unrestricted 

use in 1995. 

Sodium Disposal Facility—Also called the sodium burn pit, the Sodium Disposal Facility consisted 

of a large, rectangular, concrete-lined pit filled with water, surrounded by a concrete slab with two 

water-filled basins, a small building (Building 4886), and steam lance cleaning equipment. The 

Sodium Disposal Facility was built to clean nonradioactive metallic sodium and NaK from various 

scrap test components (pumps, valves, etc.) before the components were disposed of. In addition, the 

facility was used to treat nonradioactive waste sodium and NaK, and was also used to burn 

nonradioactive combustible liquid waste (oils, etc.). 

5.2.3 Summary of Key SSFL-Area IV Facilities 

Table 5-1 summarizes the key facilities as well as the buildings and dates of operation. 

Table 5-1: Key SSFL-Area IV Facilities, Operations, and Dates of Operation 

Table 5-1and corresponding notes span two pages. 

Operations Buildings Key Facilities Datesa

Reactor Building 4093 WBNS or L-85 Reactor: Water Boiler Neutron Source Reactor; 3 

kWthr 1956-1980 

Reactor Building 4143* SRE Reactor: Sodium Reactor Experiment Reactor; 20 MWthr 1957-1964 

SER Reactor: SNAP Experimental Reactor; 50 kWthr 1959-1960 Reactor Building 4010 S8ER Reactor: SNAP 8 Experimental Reactor; 600 kWthr 1963-1965 

S2DR Reactor: SNAP 2 Development Reactor; 50 kWthr 1961-1962 Reactor Building 4024 S10FS Reactor: SNAP 10 Flight Simulation Reactor; 7 kWthr 1965-1966 

STR Reactor: Shield Test Reactor; 50 kWthr 1961-1964 Reactor Building 4028 STIR Reactor: Shield Test and Irradiation Reactor; 1 MWthr 1964-1972 

Reactor Building 4059 S8DR Reactor: SNAP 8 Development Reactor; 600 kWthr - 1 

MWthr 1968-1969 

Reactor Building 4012 SNAP Critical Test Facility: Second SNAP Critical Test Facility 1962-1968 

Reactor Building 4373 SNAP Critical Test Facility: First SNAP Critical Test Facility 1957-1963 

Reactor Building 4019 SNAP Flight System Critical Facility 1964-1965 

OMR Critical Facility: Organic Moderated Reactor Low Power 

Critical Experiment 1958-1967 

Reactor Building 4009 SGR Critical Facility: Sodium Graphite Reactor Low Power 

Critical Experiment 1958-1967 

AETR Test Facility: Advanced Epithermal Thorium Reactor 1960-1974 

Reactor Building 4100 Fast Critical Experiment Laboratory: Used to study different 

reactor core configurations 1961-1980 

Engineering Test Building: Reactor Fuel Manufacturing Facility to 

support Sodium Reactor Experiment Nuclear 1954-1964 

Support Building 4003 

Hot Cave: Hot cell for reprocessing used reactor fuel 1954-1964 
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Table 5-1: Key SSFL-Area IV Facilities, Operations, and Dates of Operation 

Table 5-1and corresponding notes span two pages. 

Operations Buildings Key Facilities Datesa

Nuclear 

Support Building 4654 ISF: Interim Storage Facility for SRE fuels 1958-1964 

Nuclear 

Support Building 4030 Van de Graaff Accelerator: Particle accelerator 1960-1964 

Nuclear 

Support Building 4020 Hot Lab: Hot Laboratory for disassembly and examination of used reactor fuel 1957-1988 

Nuclear 

Support Building 4023 Corrosion Testing Laboratory: Liquid Metals Component Testing 

Facility 1962-1986 

Nuclear 

Support Building 4029 Radioactive Measurement Facility: Used for storage and used 

radioactive sources to calibrate radiation instruments 1959-1974 

Nuclear 

Support Building 4064 

Fuel Storage Facility: Fuel Storage Facility is a vault built to 

provide storage for fissionable fuel material (enriched uranium & 

plutonium)

1958-1993 

Nuclear 

Support 

Buildings 

4021/4022**

RMDF: Radioactive Materials Disposal Facility-built for fuel 

storage and processing solid and liquid waste for disposal in 

conjunction with the SRE operation

1959-

present 

Nuclear 

Support Building 4363 

Mechanical Component Development and Counting Building: 

Used for sodium systems in support of SRE from before 1959 until 

1963. 

1956-1963 

Notes: 

a For reactors, refers to operation of the reactor rather than program duration dates 

* Includes Component Storage Building (Building 4041), Temporary Hot Waste Storage Building (Building 4686), 

Site Service Building (Building 4163), Cold Trap Vault (Building 4695), Liquid Radioactive Waste Vault (Building 

4653), Interim Radioactive Waste Storage Area (Area 4654), and Intermediate Contaminated Storage Area (Area 

4689) 

**Includes Radioactive Material Storage Vault (Building 4022), Decontamination and Packaging Facility (Building 

4021), Low Specific Activity Waste Storage (Building 4075), Source Storage Buildings (Building 4621), Offices 

(Buildings 4034 & 4044), Non-Radioactive Material Storage (Building 4665), and Leach Field—Sanitary Sewer 

Septic Tank 

5.3 Radiological Exposure Sources from SSFL-Area IV Operations 

There were many different types of facilities and processes at Area IV of the SSFL, which included 

reactors, critical test facilities, fuel preparation and post-irradiation examination facilities, accelerator 

and calibration facilities, and support facilities. Most reactors were low-power (a few kilowatts), with 

the maximum being 20 MWt, and all had relatively short operating histories. There was a Van de 

Graaff (Deuterium-Tritium) accelerator producing neutrons with a maximum energy of 14 MeV. The 

fuel examination and manufacturing facilities, reactors, and critical facilities handled fissionable fuels 

with various enrichments, mostly compounds of uranium including carbides. They also handled 

relatively small quantities of plutonium and thorium, with the exception of Buildings 4023, 4029, and 

4030 where no thorium or plutonium was handled. Some fuel decladding took place in the hot cells of 

Area IV of the SSFL, which resulted in the release of considerable quantities of those fission products 

with high fission yields (e.g., strontium-90 and cesium-137) into the hot cell environment; also 

released were smaller amounts of europium-152, europium-154, and tritium (ORAUT-TKBS-0038-

6). The separations of irradiated fuel consisted of pilot-scale operations conducted inside a hot cell; 

therefore, exposure potential as a result of these operations was fairly small. 
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Table 5-2: Area Information and Parameters of SSFL-Area IV Nuclear Operations 

Table 5-2 and corresponding notes span three pages. 

Facility Radioisotopesa

Building 4073: 

KEWB Reactor (1 kWt) 

Building 4093: 

L-85 Reactor (3 kWt)

Building 4143: 

SRE Reactor (20 kWt)

Building 4010: 

 SER (kWt); 

S8ER (600 kWt)

Sr-90 

Cs-137 

U-234 

U-235 

U-238 

Pu-238 

Pu-239 

Pu-240 

Pu-241 

Am-241 

Building 4024: 

S2DR (50 kWt); 

S10FS (7 kWt)

Sr-90 

Cs-137 

U-234 

U-235 

U-238 

Pu-238 

Pu-239 

Pu-240 

Pu-241 

Am-241 

Co-60 

Eu-152 

Eu-154 

Th-232 

H-3 

Building 4028: 

STR (50 kWt); STIR (1 MWt)

Building 4012: 

Second SNAP Critical Test Facility

Building 4373: 

First SNAP Critical Test Facility

Sr-90 

Cs-137 

U-234 

U-235 

U-238 

Pu-238 

Pu-239 

Pu-240 

Pu-241 

Am-241 

Building 4019: 

SNAP Flight System Critical Facility 

Co-60 

Eu-152 

Eu-154 

H-3 

Building 4009: 

OMR Critical Facility; SGR Critical Facility

Building 4100: 

AETR Test Facility; 

Fast Critical Experiment Laboratory

Sr-90 

Cs-137 

U-234 

U-235 

U-238 

Pu-238 

Pu-239 

Pu-240 

Pu-241 

Am-241 
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Table 5-2: Area Information and Parameters of SSFL-Area IV Nuclear Operations 

Table 5-2 and corresponding notes span three pages. 

Facility Radioisotopesa

Building 4003: 

Engineering Test Building; 

Hot Cave

Building 4654 

 Interim Storage Facility

Building 4005: 

Uranium Carbide Fuel Pilot Plant

Sr-90 

Cs-137 

U-234 

U-235 

U-238 

Pu-238 

Pu-239 

Pu-240 

Pu-241 

Am-241 

Co-60 

Eu-152 

Eu-154 

Building 4030: 

Van de Graaff Accelerator

Sr-90 

Cs-137 

U-234 

U-235 

U-238 

Pu-238 

Pu-239 

Pu-240 

Pu-241 

Am-241 

Co-60 

Eu-152 

Eu-154 

Th-232 

Building 4020: 

Hot Laboratory

U-234 

U-235 

U-238 

Building 4023: 

Corrosion Testing Laboratory

Sr-90 

Cs-137 

U-234 

U-235 

U-238 

Pu-238 

Pu-239 

Pu-240 

Pu-241 

Am-241 

Co-60 

Eu-152 

Eu-154 

Th-232 

Building 4029: 

Radioactive Measurement Facility H-3 

Building 4064: 

Fuel Storage Facility 

Sr-90 

Cs-137 

Co-60 

Eu-152 

Eu-154 
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Table 5-2: Area Information and Parameters of SSFL-Area IV Nuclear Operations 

Table 5-2 and corresponding notes span three pages. 

Facility Radioisotopesa

Buildings 4021/4022: 

RMDF Ra-226 

Building 4363: 

Mechanical Component Development and Counting Building 

Sr-90 

Cs-137 

Source: Modified version of Table 5-9 in ORAUT-TKBS-0038-5 and Table 2-3 in ORAUT-TKBS-0038-2 

Notes: 

a Assumed particle size of 5 μm (AMAD) for all listed radionuclides 

5.3.1 Alpha 

Alpha particle emissions from radioactive source materials presented the most significant internal 

radiological protection challenge (alpha particles do not present an external exposure hazard). The 

main alpha-emitting plutonium isotopes were plutonium-238, plutonium-239, plutonium-240, and 

plutonium-241. Plutonium-238 emits 5.50 MeV (71%) and 5.46 MeV (29%) alpha particles; 

plutonium-239 emits a 5.16 MeV (73%) alpha particle with lower-abundance 5.14 MeV (15%) and 

5.11 MeV (12%) emissions. Similarly, plutonium-240 emits alphas at 5.17 MeV (73%) and at 5.12 

MeV (27%) (DOE, 2006). In addition, uranium isotopes (in varying ratios depending on whether the 

uranium being handled was natural, depleted, or enriched) were a source of alpha radiation. Uranium238 emits 4.20 MeV (77%) and 4.15 MeV (23%) alpha particles, and uranium-234 emits 4.78 MeV 

(72.5%) and 4.72 MeV (28.4%) alpha particles. There are smaller amounts of uranium-235 

(approximately 1/20 of the activity levels of uranium-238 or uranium-234 in natural uranium, varying 

up or down depending on enrichment level or depleted uranium activities, respectively) with the most 

significant alpha particles of 4.60 MeV (5%), 4.56 MeV (4.2%), 4.40 MeV (55%), 4.37 MeV (17%), 

4.22 MeV (5.7%), and others with lower energies and percent abundance (Shleien, 1992). 

5.3.2 Beta/Photon 

The fuel examination and manufacturing facilities, reactors, and critical facilities handled fissionable 

fuels with various enrichments, mostly compounds of uranium including carbides. They also handled 

relatively small quantities of plutonium and thorium. Area IV of the SSFL performed only pilot-scale 

separations of irradiated fuel; therefore, there were minimal gross fission product problems. AI did 

declad fuels in the hot cells, which resulted in the release of considerable quantities of those fission 

products with high fission yields (e.g., strontium-90 and cesium-137) into the hot cell environment. In 

addition, decladding resulted in some europium-152 and -154 releases, along with some tritium. 

The decladding was performed in a facility (hot lab) designed for such use. Strontium-90 and cesium137 are volatile and tend to migrate to the surface of the fuel, but are contained in the hot cell. The 

fuel remained contained in the hot cell until much of the short-lived activity had decayed. The facility 

was shielded and remote manipulators were used to minimize the potential for personnel exposures. 

This design reduced the external exposures from highly radioactive items to levels within 

occupational limits. A ventilation system kept the hot cells at negative pressure relative to the 

operators’ positions. Filters decontaminated the air from the hot cells prior to exhaust from the 

building (Oldenkamp, 1989). 
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Beta/photon sources consisted of reactors as well as associated radioactive material work and 

processes in buildings 4143, 4010, 4010, 4024, 4028, 4059, 4009, and 4100 that housed the reactors. 

Table 5-3 indicates the beta and photon energies and percentages at reactors and critical facilities. 

Table 5-3: Beta and Photon Energies and Percentages at Reactors and Critical Facilities 

Table 5-3 spans two pages. 

Process/ Operations 

Building Description Begin End 

Radiation 

Type 

Energy 

(keV) Percentage 

Beta >15 100 

All Reactors 1956 1980 30-250 25 Photons >250 75 

Beta >15 100 

4143 SRE Reactor 1957 1964 30-250 25 Photons >250 75 

Beta >15 100 

4010 30-250 25 SNAP Experimental Reactor and 

SNAP8 Development Reactor (S8ER) 1959 1965 Photons >250 75 

SNAP2 Development Reactor (S2DR) 1961 1962 Beta >15 100 

SNAP 10 Flight Simulation Reactor 

(S10FS3) 

4024 1965 1966 30-250 25 

SNAP Test Facilities 1971 1971 

Photons 

>250 75 

Beta >15 100 

4028 30-250 25 Shield Test Reactor and 

Shield Test and Irradiation Reactor (STIR) 1961 1972 Photons >250 75 

Beta >15 100 

4059 SNAP 8 Development Reactor (S8DR) 1968 1969 30-250 25 Photons >250 75 

Beta >15 100 

4009 30-250 25 Organic Moderated Reactor (OMR) and 

Sodium Graphite Reactor (SGR) 1958 1967 Photons >250 75 

Beta >15 100 

4100 30-250 25 Advanced Epithermal Thorium Reactor 

(AETR) 1960 1974 Photons 

(thorium) >250 75 

5.3.3 Neutron 

In uranium-235-based nuclear reactors, such as were operated at SSFL, fission neutrons have an 

average energy of about 2 MeV. In order to sustain the nuclear chain reaction, the neutrons must be 

slowed by collisions in the moderator of the reactor to an energy level of 0.025 eV (thermal neutron 

energy). Depending on the reactor design, few fast neutrons will escape the shielding. Due to the 

gamma radiation produced in the fuel and from the neutron-gamma reactions in the shielding, 

neutrons from reactors are usually a minor concern for personnel exposure compared to gamma 

radiation (ORAUT-TKBS-0038-2). 

Correspondence regarding a neutron survey report of the SRE Hot Cell containing a 14-MeV neutron 

generator lists dose rates of 75 mrem/hr fast and 11.8 mrem/hr thermal with “the assumption that all 

fast neutrons are 14 MeV” (Clow, 1966). This correspondence also states, “As further surveys are 

taken and the spectrum completely analyzed, the dose rates may be reduced” (Clow, 1966). 
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According to Clow, these surveys were performed on the roof of the facility and were higher than 

those taken at “a window” (62.2 mrem/hr fast and 7.1 mrem/hr thermal) or “at the console” (7.1 

mrem/hr fast and 2.2 mrem/hr thermal). The Van de Graaff accelerator with its D-T reaction 

generates 14-MeV neutrons that are quite monoenergetic in the 0° and 180° directions (ORAUTTKBS-0038-6, page 8). 

Table 5-4 lists facilities with neutron radiation. Facilities include reactors, accelerators, and fuel 

storage facilities. However, these facilities are assumed to be a negligible contributor to neutron doses 

due to the limited quantities of fuel present at any one time (ORAUT-TKBS-0038-6, page 8). 

Table 5-4: Facilities and Neutron Energies 

Facility Source Neutron energy (MeV) 

Reactors Reactors 

Plutonium Fuel Storage Building 

4064 

Spontaneous fission and alpha-neutron 

reactions 

0.1-2.0, WR

*

=20 

4030 Van de Graaff Accelerator D-T reaction 2-14 max, WR

*

=10 

Notes: 

* The distribution of energies and International Commission on Radiological Protection Publication 60 radiation 

weighting factors

5.3.4 Incidents and Fires 

The most radiologically significant and notable incident at Area IV of the SSFL was associated with 

the SRE and occurred in July 1959. The SRE reactor underwent an accidental partial blockage of 

sodium coolant flow in some of the reactor coolant channels. This resulted in the partial melting of 13 

reactor fuel assemblies and the release of some fission products, which contaminated the reactor 

cooling system. All of the reactor safety systems functioned, and the reactor was shut down. The 

reactor fuel assemblies were removed, inspected, and stored at the Radioactive Materials Disposal 

Facility (RMDF). They were later declad in the Hot Lab, and the fuel and cladding were shipped 

offsite (ORAUT-TKBS-0038-2). 

Based on its research, NIOSH has identified other incidents with minimal (or no) radiological 

impacts; these other incidents are included in summary form in Nuclear Operations at Rockwell’s 

Santa Susana Field Laboratory – A Factual Perspective as well as in various site incident reports and 

documents collected by NIOSH (Oldenkamp, 1989). NIOSH obtained additional clarification 

regarding these types of incidents and radiological occurrences during discussions with former and 

current SSFL employees (Personal Communication with Fireman; Personal Communication with 

Health Physicist-November 15, 2007; Personal Communication with Health Physicist-November 20, 

2007; Personal Communication with HP-Radiation Engineer; Personal Communication with Health 

and Safety Manager). While additional information regarding these incidents and occurrences has not 

been included in this report, it is available in the identified references and the SSFL Site Profile 

documents. 
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6.0 Summary of Available Monitoring Data for the Class Evaluated 

by NIOSH 

NIOSH has located personal and area monitoring data in the NIOSH Site Research Database (SRDB) 

and in NOCTS which have been used to estimate doses to individual employees in the proposed class. 

The results of NIOSH data reviews are provided in the following subsections. 

6.1 SSFL-Area IV Internal Monitoring Data 

Radionuclides of concern for the facilities and process at Area IV of the SSFL were primarily fission 

products. Exposure to fissionable material, transuranics, and activation products was also possible. 

Summaries of the available in vitro and in vivo data, as well as general overviews of sampling and 

analytical protocols, are provided in the following subsections. Additional details, including 

analytical methods, detectable activities, and reporting protocols can be found in ORAUT-TKBS0038-5. 

6.1.1 Internal Urinalysis Data 

NIOSH has identified only limited amounts of internal personnel monitoring data for pre-1959 

exposures; this is consistent with its findings that a SSFL routine bioassay program was not initiated 

until August 1958 (Kellehar, 1966). Therefore, NIOSH has concluded that pre-1959 workers will 

generally not have bioassay data prior to August 1958 (except in rare cases). After August 1958, 

routine bioassay was normally in the form of urinalysis at a three-month frequency. At first, gross 

alpha or gross beta measurements were made of the samples. Specific radionuclides could be 

determined “where required” by performing additional radiochemical separations and analyses (Lang, 

1960). For the period 1959 to 1966, analyses were available for uranium (both fluorometric and 

radiometric), mixed fission products, polonium-210, plutonium, strontium-90, tritium, and thorium 

(ORAUT-TKBS-0038-5). 

Table 6-1 lists the frequencies for in vitro monitoring. 

Table 6-1: Internal Dose Control Program (in vitro) 

Routine Monitoring Type Period Frequency 

Urine, single voida 1958-1988 Quarterly, semiannual, annual, based on job 

Urine, single void, tritium 1958-1966 Weekly, based on job 

Notes: 

a

 Sample requested on Friday, first voiding on Monday morning requested. Positive samples were verified and 

followed with at least one 24-hr sample collected on Sunday. 

Table 6-2 lists the minimum detectable activities (MDAs) and reporting levels for periods 

corresponding to the bioassay methods discussed in this section. The reporting levels are listed in the 

units quoted in the references, which are generally the units of the results. However, various volumes 

were used to report the results. In general, the excretions assumed when reporting “per sample” or 

“per day” were 1,500 mL for urine samples and 135 g for fecal samples. 
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Table 6-2: Detection limits 1958–1966a 

Table 6-2 and corresponding notes span two pages. 

Radionuclide Method/ 

description MDAb Reporting 

levelc

Gross alpha Urine 15 dpm/L (U. S. Nuclear) 

4 dpm/sample (Nuclear Science and Engineering Corporation) 

7.5 dpm/L 

2 dpm/sample 

Gross beta Urine 150 dpm/L (U. S. Nuclear) 

2 dpm/mL (Nuclear Science and Engineering Corporation) 

75 dpm/L 

1 dpm/mL 

Gross beta 

(minus K-40) Urine 0.4 dpm/mL (BioScience) 

0.04 dpm/mL (U.S. Testing Inc.) 

0.2 dpm/mL 

0.02 dpm/mL 

Tritium Urine 

4,440 dpm/mL (Nuclear Science and Engineering Corporation) 

10,000 dpm/mL (BioScience) 

10,000 dpm/mL (U.S. Testing Inc.) 

2,220 dpm/mL 

5,000 dpm/mL 

5,000 dpm/mL 

MFP (gross) Urine 

4 dpm/mL (Nuclear Science and Engineering Corporation) 

60 dpm/sample (Hanford) 

0.2 dpm/mL (BioScience) 

0.04 dpm/mL (U.S. Testing Inc.) 

2 dpm/mL 

30 dpm/sample 

0.1 dpm/mL 

0.02 dpm/mL 

Polonium-210 Urine 0.02 dpm/mL (BioScience) 

0.02 dpm/mL (U.S. Testing Inc.) 

0.01 dpm/mL 

0.01 dpm/mL 

Plutonium Urine 0.0006 dpm/mL (U.S. Testing Inc.) 0.0003 dpm/mL 

Strontium-90 Urine 0.2 dpm/mL (BioScience) 

0.04 dpm/mL (U.S. Testing Inc.) 

0.1 dpm/mL 

0.02 dpm/mL 

Thorium Urine 

0.4 μg/125 mL (Tracerlab) 

0.0004 μg/mL (BioScience) 

0.0004 μg/mL (U.S. Testing Inc.) 

0.2 μg/125 mL 

0.0002 μg/mL 

0.0002 μg/mL 

Uranium UF-1A 

Urine 

0.004 μg/mL (U. S. Nuclear) 

0.0002 μg/mL (Controls for Rad.) 

0.0004 μg/mL (Nuclear Science and Engineering Corporation) 

0.0004 μg/mL (BioScience) 

0.0004 μg/mL (U.S. Testing Inc.) 

0.002 μg/mL 

0.0001 μg/mL 

0.0002 μg/mL 

0.0002 μg/mL 

0.0002 μg/mL 

Uranium (enriched) UR-1B 

Urine 

15 dpm/L (U. S. Nuclear) 

4 dpm/sample (Nuclear Science and Engineering Corporation) 

0.012 dpm/mL (BioScience) 

0.012 dpm/mL (U.S. Testing Inc.) 

7.5 dpm/L 

2 dpm/sample 

0.006 dpm/mL 

0.006 dpm/mL 

Notes: 

a

 The date of the reference for the reporting levels, which are the best information available for the period 1958-1966. 

b

 Assumed to be twice the sensitivity (see ORAUT-TKBS-0038-5 text). The reported sensitivity (“less than” values) in the 

claimant records should be used to determine MDA in lieu of the values listed in this table. These values should also be used 

for the in-house laboratory. c Sheperd, 1959; O’Brien, 1959; Edelmann, 1959; and Lee, 1963. UST results that were considered to be less than detectable 

generally have the reporting level included with the result. 

NIOSH has access to urinalysis results and other supporting data for the period after the initiation of 

the routine bioassay program (as the routine bioassay program started in August 1958 and the initial 

collection frequency was three months, this corresponds to the end of 1958). The practice at the 

facility was to collect urine samples based on job assignments that required exposures to radioactive 

materials (Kopenhaver, 1986; Garcia, July 1, 1963; Alexander, 1959). Details regarding the various 

analyses used, and the associated minimum detectable activities, are presented in ORAUT-TKBS0038-5. 
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6.1.2 Lung, Whole-Body, and Other Types of Internal or Bioassay Data 

Routine lung-counting, using equipment and techniques specifically developed to measure lung 

deposits of uranium, did not start until 1968 (outside of the timeframe evaluated in this report). 

Whole-body counts for fission or activation products occurred sporadically from 1968 through 1983 

(also outside of the timeframe evaluated in this report). Although there is an indication that fecal 

samples were included in the bioassay program in 1970, no evidence has been recovered of their use 

during the period being evaluated in this report. There is evidence that wound contamination was 

evaluated. However, because urinalysis data are available, further review and evaluation of wound 

contamination data are not being pursued in this report. Finally, SSFL was neither a storage nor a 

processing location for large quantities of uranium or thorium; most of this material was in the form of 

reactor fuel that had been processed elsewhere to remove any long-lived radon- or thoron-producing 

progeny. 

Based on the timeframe of the monitoring as compared to the proposed worker class time period 

evaluated in this report, these data were not evaluated for the dose reconstruction feasibility 

evaluation. Therefore, further discussion or evaluation of lung or whole-body counts (in vivo), fecal 

sampling, wound contamination evaluations, or evaluations of radon or thoron exposures is 

considered outside the scope of this SEC petition evaluation. 

6.2 SSFL-Area IV External Monitoring Data 

The external dose received by workers was a function of the physical location of the workers on the 

site, the processes taking place, the types and quantities of material present, and the time spent in each 

location. The radiation contributing to external dose was primarily beta, photon (including X-rays) 

and neutron. Information associated with the radiological external monitoring program at Area IV of 

the SSFL available to NIOSH has been reviewed as a part of this SEC (Badger, 1965-1987; Various 

authors, 1970-1987). 

AI used a two-element dosimeter to measure and account for the photon and beta exposures in the 

early years (1954-1962) at Area IV of the SSFL; the two-element dosimeter was very similar to the 

film dosimeters used at the Y-12 plant (ORAUT-TKBS-0014-6) and the Hanford Site (ORAUTTKBS-0006-6) that were based on the design developed by Pardue, Goldstein, and Wollan (Pardue, 

1944). In addition to the two-element film badges, the SSFL also incorporated the use of pocket 

ionization chambers (PICs or pencil dosimeters). In 1963, AI started using commercial vendors 

(Landauer) who supplied a multi-element dosimeter to measure and account for photon and beta 

exposures; this practice continued throughout the rest of the facility’s operating life (ORAUT-TKBS0038-6). On occasion, workers wore special dosimeters to monitor non-routine work such as “hot 

jobs” or cell entries. NIOSH has not determined whether the special dosimeter procedure was in 

operation before 1963 (ORAUT-TKBS-0038-6). 

In terms of dosimeter exchanges at SSFL, Lang (1960) states, “Normally film badges are to be worn 

for monthly periods except where the possibility of exceeding 100 mrem/week of exposure is 

expected. Then the badges are to be analyzed more frequently.” Starting in 1963, most dosimeters 

were processed by the vendor (Garcia, April 5, 1963). Personnel doses from Landauer before 1985 

were reported in milliroentgen as either penetrating (photon) or non-penetrating (beta) exposure. The 

reported total dose values included any special dosimeter results sent by AI for that exchange period. 
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AI used Nuclear Track Emulsion, Type A (NTA) film to measure neutron exposures over the history 

of operations at Area IV of the SSFL (beginning with the start of reactor operations and the use of the 

Van de Graaff accelerators). Both fast and thermal neutrons were measured and recorded as wholebody dose in rem. The routine exchange frequency (biweekly or monthly) and evaluation of the 

neutron dose is described further in the SSFL site program data, procedures, and in ORAUT-TKBS0038-6 (Badger, 1965-1987; Various authors, 1970-1987). NIOSH has also identified and located 

area neutron surveys associated with the operations where potential neutron exposures existed, which 

can be used to supplement and evaluate the personnel external neutron monitoring data (ORAUTTKBS-0038-6). Table 6-3 lists the time periods, types of dosimeters, exchange periods, MDLs, and 

estimated missed doses for the AI dosimeters used at Area IV of the SSFL. 

Table 6-3: Estimated Exchange Frequencies, MDLs, and Annual Missed Doses 

Radiation 

Type 

Time 

Perioda Dosimeter Exchange Period 

MDL 

(rem)b

Estimated 

Annual Missed 

Dose (rem)c

Daily 0.005 0.625 

Weekly 0.04 1.04 

Bi-weekly 0.04 0.52 1954-1962 Site-specific two element film 

PIC 

Monthly 0.04 0.24 

Monthly 0.04 0.24 1963-1979 Landauer multi-element film Quarterly 0.04 0.08 

Photon - Beta 

1980d

-Present Landauer multi-element film Quarterly 0.01e 0.02 

Bi-weekly <0.05 0.650 Neutron 1954-Present NTA film Monthly <0.05 0.30 

Notes: 

a The pre-Landauer start dates for the dosimeter time period are estimated based on the earliest entries in the 

available records. Landauer or its predecessor was the vendor in 1963 (Garcia, April 5, 1963). 

b Estimated MDLs for each dosimeter in the workplace, although many doses were reported at less 

 than the MDL. 

c

 Estimated annual missed dose calculated using MDL/2 from OCAS-IG-001. 

d Neutron dosimeters continued monthly exchanges (Badger, 1965-1987; Various authors, 1970-1987). 

e

 Yoder, 2005. 

NIOSH has access to photon, beta, and neutron external dosimetry results, as well as other supporting 

data for the entire period evaluated in this report (available for all years of site operation). The policy 

at SSFL was to assign the applicable dosimetry to anyone with the potential for photon, beta, or 

neutron exposure; it was assigned based on job assignments that required exposure to radioactive 

materials (Badger, 1965-1987; Various authors, 1970-1987; ORAUT-TKBS-0038-6). Details 

regarding the various analyses used, and the associated minimum detectable activities, are presented 

in ORAUT-TKBS-0038-6. 

6.3 SSFL-Area IV Air Sampling Data 

The SSFL Area IV health physics program records include a number of air samples for alpha- (in 

limited cases) and beta-emitters. Measured air concentrations can be useful for estimating maximum 

doses, when bioassay data are unavailable either because the worker was unmonitored or in cases of 

non-radiological workers (both cases are associated with individuals with lower exposure potentials as 

compared to the radiological workers at Area IV of the SSFL). If needed, air concentration 
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measurements are used to assign internal doses from radioactive material in the general environment, 

in accordance with ORAUT-TKBS-0038-4. 

An environmental monitoring program, including air emissions and ambient air monitoring, was 

established at Area IV of the SSFL in May 1954, prior to completion of the first radiological facility 

(Sapere Consulting, 2005). From 1955 to present, ambient gross beta activity in air has been 

measured continuously in Area IV of the SSFL, but there are very few, if any, radionuclide-specific 

data for the years being evaluated in this report. In all cases, the gross beta concentrations were 

attributed to natural background plus fallout from aboveground nuclear weapons tests. Because of 

this, potential worker environmental inhalation intakes from onsite atmospheric radionuclide 

concentrations can be estimated using average annual concentrations of facility stack emissions 

(ORAUT-TKBS-0038-4). Specific radionuclides released from various facilities in stack emissions 

are identified in data from 1988 through 1999; these data were used to characterize radionuclide 

emissions for all years (ORAUT-TKBS-0038-4). The radionuclide-specific data showed that 

naturally-occurring radionuclides were major constituents of radionuclide emissions, with polonium210 a main constituent of the gross alpha concentration and potassium-40 and beryllium-7 major 

constituents of the gross beta detected. Any uranium and thorium detected is assumed to originate 

from facility activities (albeit it could be naturally occurring). 

The available stack and air monitoring data at SSFL-Area IV have not been found to be sufficiently 

process-related or facility-specific to support the evaluation of pre-1959 internal doses where other 

internal monitoring data are lacking. Because internal monitoring is available for the period after 

1958, further discussion or evaluation of air sampling will not be included in this SEC petition 

evaluation. 

7.0 Feasibility of Dose Reconstruction for the Class Evaluated by 

NIOSH 

The feasibility determination for the proposed class of employees covered by this evaluation report is 

governed by both EEOICPA and 42 C.F.R. § 83.13(c)(1). Under that Act and rule, NIOSH must 

establish whether or not it has access to sufficient information either to estimate the maximum 

radiation dose for every type of cancer for which radiation doses are reconstructed that could have 

been incurred under plausible circumstances by any member of the class, or to estimate the radiation 

doses to members of the class more precisely than a maximum dose estimate. If NIOSH has access to 

sufficient information for either case, NIOSH would then determine that it would be feasible to 

conduct dose reconstructions. 

In determining feasibility, NIOSH begins by evaluating whether current or completed NIOSH dose 

reconstructions demonstrate the feasibility of estimating with sufficient accuracy the potential 

radiation exposures of the class. If the conclusion is one of infeasibility, NIOSH systematically 

evaluates the sufficiency of different types of monitoring data, process and source or source term data, 

which together or individually might ensure that NIOSH can estimate either the maximum doses that 

members of the class might have incurred, or more precise quantities that reflect the variability of 

exposures experienced by groups or individual members of the class. This approach is discussed in 

OCAS’s SEC Petition Evaluation Internal Procedures, which are available at 

http://www.cdc.gov/niosh/ocas. The next four major subsections of this Evaluation Report examine: 
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• The sufficiency and reliability of the available data. (Section 7.1) 

• The feasibility of reconstructing internal radiation doses. (Section 7.2) 

• The feasibility of reconstructing external radiation doses. (Section 7.3) 

• The bases for petition SEC-00093 as submitted by the petitioner. (Section 7.4) 

7.1 Pedigree of SSFL-Area IV Data 

This subsection answers questions that need to be asked before performing a feasibility evaluation. 

Data Pedigree addresses the background, history, and origin of the data. It requires looking at site 

methodologies that may have changed over time; primary versus secondary data sources and whether 

they match; and whether data are internally consistent. All these issues form the bedrock of the 

researcher’s confidence and later conclusions about the data’s quality, credibility, reliability, 

representativeness, and sufficiency for determining the feasibility of dose reconstruction. The 

feasibility evaluation presupposes that data pedigree issues have been settled. 

A centralized radiological records program was established at SSFL at the beginning of nuclear 

operations in 1955. The radiological safety program began with the first proposal to secure 

radioactive material and continued to expand with the level of operations. All health physics records 

have been maintained in a centralized location at SSFL-Area IV and are readily available to NIOSH. 

They consist of personnel monitoring records (external exposure and bioassay), meteorological 

monitoring, environmental monitoring, and effluent monitoring records (Lang, 1960; Various authors, 

1960-1966). The availability and extent of the health physics records was corroborated in November, 

2007 in a telephone interview with the SSFL Manager of Radiation Safety and in a subsequent data 

capture trip to the site to do additional research on the data in the available files. Since revision 1 of 

this evaluation report, 458 documents have been added to the NIOSH Site Research Data Base as a 

result of additional data capture efforts. 

Examination of the internal and external dosimetry data available for SSFL employees indicates that 

the data are of sufficient accuracy and represent the range of exposures associated with the work 

performed during the covered period. The type of monitoring was well-suited for the type of work 

performed at the facility and was comparable to what was used by other facilities doing similar work 

at that time. 

In addition to health physics records, the available source term and process information includes the 

types and quantities of specific radionuclides and sources present, their chemical and physical form, 

and the types and frequency of operations (including controls) in which these sources were utilized. 

NIOSH has determined that there is a lack of meaningful source term information (such as isotopic 

and curie content) for the 1955-1958 period, which limits the development of a reliable source term 

for the SSFL-Area IV. However, sufficient information is available to support the evaluation of dose 

for the post-1958 employment periods using internal and external monitoring data. 
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7.1.1 Internal Monitoring Data Review 

The routine in vitro bioassay program was established in August 1958 (Kellehar, 1966). The practice 

at SSFL was to collect urine samples based on job assignments that required exposure to radioactive 

materials. Prior to August 1958, NIOSH has found only limited personnel bioassay data or associated 

internal monitoring-related information for the worker class being evaluated in this report. However, 

NIOSH does have access to original urinalysis results and other supporting program data for the 

period after the initiation of a bioassay program in August 1958. 

NIOSH reviewed the internal monitoring data analyses performed as part of a mortality study as well 

as a comprehensive dose reconstruction methodology study at Rocketdyne/Atomics International 

(Boice, 2006; Boice, May 2006). These documents and studies demonstrate the availability of the 

internal monitoring data collected over the operational period evaluated in this report and that the site 

collected and appropriately analyzed samples from individuals with the potential for internal 

radiological exposures. In addition, these documents include data for the highest exposed individuals 

for post-1958 periods, and the data are of sufficient quality and quantity to support radiological dose 

reconstruction and/or coworker studies/analyses for the post-1958 period (Alexander, 1959; Boice, 

2006; Garcia, July 1, 1963; Garcia, July 15, 1963; Boice, May 2006; Kellehar, 1966; Lang, 1960; Lee, 

1963; O’Brien, 1959; Sheperd, 1959; Various Authors, 1960-1966). 

Throughout the history of the site, bioassay samples were also analyzed by commercial laboratories. 

These commercial laboratories incorporated the same practices and procedures required by SSFL 

(Rocketdyne/Atomics International) bioassay programs and generally incorporated state-of-the-art 

equipment and analysis techniques, kept abreast of the latest AEC/DOE requirements and regulations, 

and incorporated the latest quality control and assurance methods (which would be expected as 

internal monitoring was their business and would support retaining additional AEC/DOE-related 

business). During the large-scale studies by Boice (2006, May 2006), 200,000 records in 14,189 

worker folders were reviewed, and 2,232 workers were determined to have been internally monitored. 

A total of 27,023 records of internal monitoring data were compiled (Boice, 2006; Boice, May 2006). 

Based on this information, NIOSH has concluded that SSFL policies for the collection and 

maintenance of employee monitoring data are sufficient for dose reconstruction in terms of the 

pedigree parameters described previously. As discussed with former and current employees from the 

site, and later confirmed during data captures (including the most recent data capture in December 

2007), physical records are available in hardcopy form and are stored onsite in file cabinets. Upon 

request, the SSFL provides original hardcopy records for individual claimants; these records have 

been used by NIOSH and support NIOSH’s ability to confirm data. The lack of available internal 

monitoring data for the pre-1959 period prevents the performance of an internal consistency and data 

pedigree check for the Area IV SSFL data for the period from 1955 through 1958. Having the 

original Area IV records of SSFL personnel and the associated radiological program information 

eliminates the need for a separate internal consistency check of the internal data for the period from 

1959 through 1965. 

7.1.2 External Monitoring Data Review 

When operations began at Area IV of SSFL, the site used its own dosimeter to monitor photon and 

beta exposures. The dosimeter design was very similar to the film dosimeters used at the Y-12 plant 

and the Hanford site, which were based on the design developed by Pardue, Goldstein, and Wollan 
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(Pardue, 1944). During this time period, AI supplemented their film dosimetry with pencil dosimeters 

and NTA film that was used to monitor neutron exposures. Both penetrating and non-penetrating 

doses were documented in personnel records (ORAUT-TKBS-0038-6). 

AI started using commercial vendors for its dosimetry needs in the early 1960s and continued that 

practice to the end of its operating life. Original individual external dosimetry records are available 

for SSFL workers, and all workers shown to have a potential for radiation exposure were monitored 

over the period evaluated in this report (and for the entire operational period of Area IV of the SSFL). 

The exchange frequencies for personnel at the site were based on the potential to exceed the 

administrative exposures limits for the applicable period in the radiological program (Lang, 1960; 

ORAUT-TKBS-0038-6). Having the original personnel records and the associated radiological 

program information for Area IV of SSFL eliminates the need for a further consistency check of the 

external data for the period from 1955 through 1965. 

As discussed in a mortality study and a comprehensive dose reconstruction methodology study at 

Rocketdyne/Atomics International (Boice, 2006; Boice, May 2006), radiation workers (workers 

routinely assigned to areas where exposure was likely) were assigned a dosimeter. During the early 

years of the external monitoring program, all employees were assigned a radiation folder (to store any 

radiological monitoring data collected for each individual) regardless of whether or not they were 

designated as radiation workers (which resulted in the identification of a multitude of empty radiation 

folders during the studies). Based on the reviews performed as part of the studies, it was confirmed 

that the radiation workers who had the potential to receive routine external radiological exposures 

were monitored using dosimeters for all years evaluated in this report (and for all years of site 

operations). Therefore, NIOSH has concluded that the available information includes data for the 

highest exposed individuals for the period evaluated in this report (and for the operations of the site), 

and that these data are of sufficient quality and quantity to support radiological dose reconstruction 

and/or coworker studies/analyses for the entire evaluated period. As discussed in the dose 

reconstruction methodology study and the available information describing the external monitoring 

and dosimetry program during the period evaluated in this report, sufficient information exists 

(including quality control information and data to support uncertainty analyses) to support interpreting 

the external monitoring data for the entire period evaluated in this report (Badger, 1965-1987; Boice, 

2006; Clow, 1966; DOE, 1986; Garcia, April 5, 1963; Garcia, July 1, 1963; Garcia, July 15, 1963; 

Boice, May 2006; Lang, 1960; Pardue, 1944; Various Authors, 1970-1987). Boice (2006, May 2006) 

reviewed over 200,000 records in 14,189 worker folders, and determined that 5,743 workers were 

externally monitored (Boice, 2006; Boice, May 2006). 

7.2 Internal Radiation Doses at SSFL-Area IV 

The principal sources of internal radiation doses for members of the evaluated class were airborne 

alpha and beta-gamma particulate materials. Alpha-emitting radionuclides were associated with 

fissionable materials and transuranics in reactor fuel. They include plutonium-238, plutonium-239, 

uranium-234, uranium-235, uranium-238, thorium-228, thorium-230, thorium-232, and americium241. Reactor operations produced beta-gamma-emitting mixed fission products and activation 

products. These include cobalt-60, strontium-90, cesium-137, and tritium. Workers were potentially 

exposed to these radionuclides during reactor operations as well as during shutdowns, modifications, 

and refueling. Dust containing radionuclides could be inhaled by individuals and then deposited in 

the respiratory tract. The dust would also settle on surfaces and become resuspended in the air, where 

it could again be inhaled or ingested by transfer from contaminated surfaces via hand-to-mouth. The 
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majority of the exposures over the history of SSFL-Area IV operations are considered to be chronic in 

nature. There were a number of incidents that could have resulted in higher short-term exposures for a 

limited number of employees. The incidents are well-documented in the site records (ORAUTTKBS-0038-5). 

In addition to particulate matter, reactor operations also produced some short-lived gaseous (xenon133 and krypton-85) and volatile fission products (radioiodines). Workers could have inhaled this 

material, although generally there were systems designed to confine or absorb this material until it 

decayed or was disposed of as waste (Sapere Consulting, 2005). 

7.2.1 Process-Related Internal Doses at SSFL-Area IV 

The following subsections summarize the extent and limitations of information available for 

reconstructing the process-related internal doses of members of the evaluated class. Most of the 

information summarized below is provided in greater detail in the individual TBDs and other 

document sources. 

7.2.1.1 Urinalysis Information and Available Data 

As discussed in Section 6.1, NIOSH has identified only limited amounts of internal personnel 

monitoring data for pre-1959 exposures. The routine in vitro bioassay program was established in 

August 1958 (Kellehar, 1966). NIOSH has access to urinalysis results and other data for the post1958 period evaluated in this report. Quarterly urinalysis was normally used for routine bioassay 

during the early years of the program. Initially, gross alpha or gross beta measurements were 

analyzed and specific radionuclides could be determined “where required.” Should significant 

activity be detected, more detailed radiometric analysis could be performed with suitable body-burden 

determinations (Lang, 1960). For the period of 1958 through 1966, analyses were performed for 

uranium (both fluorometric and radiometric), mixed fission products, plutonium, strontium-90, 

tritium, and thorium (Alexander, 1959). In addition, NIOSH has access to the internal monitoring 

program data and radiological source term information associated with all radiological operations 

performed during (and after) this time period. For the operational period prior to 1959, NIOSH only 

has access to limited internal monitoring data, and lacks access to sufficient program data or a 

developed source term evaluation. 

The urinalysis practice at SSFL-Area IV for the period after 1958 included collecting urine samples 

based on job assignments that required exposure to radioactive materials. Workers who were not 

monitored by urinalysis were considered to be at little or no risk of exposure. Special urine samples 

were requested when an intake was suspected, or if a routine sample exceeded 10% of the urinary 

excretion expected from a single maximum permissible body burden. Special samples were analyzed 

by the onsite laboratory, commercial labs, or both (Kellehar, 1966). 

Based on limited pre-1959 internal monitoring data and associated program or source term 

information, NIOSH has concluded that there is insufficient information to support establishing a 

bounding internal exposure scenario for the proposed worker class that worked in Area IV of the 

SSFL from 1955 through 1958. NIOSH has identified post-1958 radiological internal monitoring 

program data and has access to the original monitoring data for individuals working in the potential 

exposure areas at Area IV of the SSFL. Therefore, NIOSH has concluded that there are sufficient data 
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to bound the dose (reconstruct dose with sufficient accuracy) for the remainder of the evaluated 

worker class for the period from 1959 through 1965. 

7.2.2 Ambient Environmental Internal Radiation Doses at SSFL-Area IV 

Ambient environmental air sampling data exist for the period evaluated in this report; however, no 

stack effluent concentration data are available for the period prior to 1965. The facilities operating 

during the pre-1965 period were mainly small reactor facilities. The SRE, the largest reactor at SSFL, 

operated from 1957 (full-power by May 1958) to 1964 in Area IV. 

 

The SSFL ambient air monitoring data through 1999 are reported as gross alpha and gross beta 

without identification of specific radionuclides. In all cases, the gross alpha and gross beta 

concentrations are a small fraction of the applicable limits and are indistinguishable from background. 

Because of this, stack emission data were used as the primary basis for inhalation intake estimates 

from onsite atmospheric radionuclide concentrations. Furthermore, the average percentage that each 

radionuclide contributed to the gross alpha or gross beta concentration was determined from 1988-to1999 data that were applied to earlier years to estimate radionuclide-specific stack concentrations. 

Any uranium and thorium detected was assumed to originate from facility activities, even though it 

could be naturally occurring (ORAUT-TKBS-0038-4). 

The facility effluent concentrations for each year were summed and worker exposure hours and rates 

were estimated (ICRP, 1975). Because the stack effluent concentrations were at the point of release, a 

further reduction factor of 0.01 was applied to account for the decreased overall intake due to 

contributions from facility spacing, atmospheric dispersion, and building wake effects (ORAUTTKBS-0038-4). 

Noble gases were not included in inhalation estimates because they were not included in emission 

monitoring data. The potential dose from radioactive noble gases is very small. Acute release events 

that would contribute negligibly to the annual average radionuclide concentration are not included in 

the annual average effluent concentrations (ORAUT-TKBS-0038-4). 

The use of facility ambient air monitoring data for assessing internal radiation doses was evaluated for 

performing EEOICPA dose reconstructions. However, NIOSH has concluded that the ambient 

environmental method (described above) does not support a bounding internal exposure evaluation 

method for the pre-1959 time period, when the facility lacked a routine urinalysis program. These 

data could be applied to support partial internal dose reconstructions (for non-presumptive cancers in 

the pre-1959 era or cases in the pre-1959 era with less than 250 days employment). For the post-1958 

timeframe, personnel ambient exposure/dose is accounted for within the available urinalysis 

monitoring data for those personnel, and these available bioassay data can be used to bound the post1958 ambient/environmental doses. Based on this information, further evaluation of these exposures 

was not performed or required for this evaluation report. 

7.2.3 Internal Dose Reconstruction 

When Area IV was established at SSFL in 1953, some existing structures were converted for use in 

nuclear research and development, but many new buildings had to be specifically constructed to house 

the small reactors and critical assemblies that were the main focus of the early years. Many of the 

early facilities were not operational until after 1955 (Sapere Consulting, 2005). Small reactors and 
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critical assemblies could produce high external doses in unshielded areas and could have presented 

some possibility of exposure to enriched uranium; however, in general these did not operate at high 

enough power levels to produce large quantities of fission products or transuranics (Sapere 

Consulting, 2005). This is consistent with the lack of a bioassay program during this time at Area IV 

of the SSFL and the logic for initiation of a bioassay program at the site in August 1958 (as operations 

were ramping up). 

It is important to note that the 1959 SRE incident occurred after the start of routine bioassay and that 

some of the highest exposure-potential facilities, such as the Hot Laboratory (Building 4020), the 

Nuclear Materials Development Facility (Building 4055), and Uranium Carbide Fuel Pilot Plant 

(Building 4005), were not active until after the start of the routine bioassay program (Sapere 

Consulting, 2005). 

The following subsections summarize the methods available to dose reconstructors to either estimate 

the maximum internal radiation dose or estimate the internal radiation doses to members of the class 

more precisely than a maximum dose estimate. 

Workers employed prior to 1959—Radionuclides of concern were primarily fission products and 

uranium. Some exposure to fissionable material, transuranics, and activation products was also 

possible (Sapere Consulting, 2005). Since the routine bioassay program was not initiated until August 

1958, many of these workers will not have bioassay data prior to that date. In the rare case that pre1959 bioassay results do exist for an individual, this data may be applied to support a partial internal 

dose reconstruction if there is information available to support interpreting the data and/or results. For 

workers without bioassay results, a partial reconstruction or estimation of their dose is possible 

through evaluation of their post-1958 internal monitoring data (if available) and/or application of the 

ambient environmental dose estimates. 

Workers employed from 1959-1965—Radionuclides of concern were primarily fission products. 

Exposure to fissionable material, transuranics, and activation products became more likely as the 

number, variety, and power levels of the small reactors increased and new facilities for handling 

unsealed radioactive materials came online (Sapere Consulting, 2005). Doses to the organ of interest 

for the majority of these workers can be calculated from bioassay results. Although some early 

bioassay results are recorded only in terms of gross activity, ORAUT-OTIB-0054 provides guidance 

on the assignment of radionuclide-specific intakes of mixed fission and activation products when air 

sampling or urinalysis data associated with reactors or reactor fuels are available only as gross or total 

beta activity or gross or total gamma activity. The availability of actual bioassay results (and 

associated program data and documentation/information that support the ability to interpret the 

internal data) for members of the evaluated worker class who were considered radiation workers (and 

therefore have post-1958 bioassay results available), supports NIOSH’s ability to reconstruct dose for 

those members of the evaluated class with sufficient accuracy. Because these data would include the 

worst-case exposure scenarios for the evaluated worker class, as discussed in Section 7.1.1, NIOSH 

contends that the availability of these data also supports its ability to establish a bounding exposure 

scenario for the entire evaluated worker class. Although research by NIOSH has indicated that those 

radiation workers who had an internal exposure potential were monitored, NIOSH has deemed it 

useful to investigate the feasibility of an internal coworker model based on the entirety of the 

abstracted internal dose records collected by Boice et al. (2006, May 2006). Data from the coworker 

model could be applied in cases where a worker was not monitored, but by today’s standards should 
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have been monitored for internal exposure or for a worker for whom the internal data are incomplete 

or were lost. 

7.2.4 Internal Dose Reconstruction Feasibility Conclusion 

NIOSH has established that unmonitored intakes of radionuclides associated with reactor operations 

(and associated operations) may have occurred in Area IV of the SSFL prior to 1959 and that these 

exposures may have resulted in unmonitored doses to site workers and workers incidentally exposed 

in these facilities. NIOSH has also determined that there was no established routine bioassay program 

for Area IV of the SSFL prior to August 1958. There is insufficient information associated with the 

source terms for the operations during this timeframe to support developing a method to estimate or 

bound internal radiological exposures based on source term (or any currently available) information or 

data. For this reason, NIOSH has concluded that it cannot bound or reconstruct pre-1959 internal 

doses with sufficient accuracy for any member of the evaluated worker class. NIOSH has established 

that it can provide a partial estimate of dose for the pre-1959 workers (for non-presumptive cases or 

for < 250-work day cases in the time period) using any available data in the period, the available 

monitoring data after 1958 (back-extrapolating the potential intake/dose), or ambient environmental 

dose reconstruction information. 

For the period after 1958 (from 1959-1965), NIOSH has established that it has access to sufficient 

internal dose information to either: (1) estimate the maximum internal radiation dose for every type of 

cancer for which radiation doses are reconstructed that could have been incurred under plausible 

circumstances by any member of the class; or (2) estimate the internal radiation doses to members of 

the class more precisely than a maximum dose estimate. 

7.3 External Radiation Doses at SSFL-Area IV 

There were many different types of facilities and processes at Area IV of the SSFL with sources of 

external exposure, including reactors, critical test facilities, fuel preparation and post-irradiation 

examination facilities, accelerator and calibration facilities, and support facilities. The major 

accelerator was a Van de Graaff generator that produced neutrons. The fuel facilities, reactors, and 

critical facilities handled fissionable fuels with various enrichments, mostly compounds of uranium 

(including carbides). They also handled relatively small quantities of plutonium and thorium. Smallscale fuel decladding operations also resulted in releases of fission products and tritium within the hot 

cell environment (ORAUT-TKBS-0038-6) 

The principal source of external radiation doses for members of the evaluated class included exposure 

to beta particles, gamma rays, and neutrons. The radiation sources contributing to these exposures 

have been identified as: 

• Operations involving the reactor research and nuclear operations, 

• Operations involving nuclear support (fuel, disassembly and examination of reactors, fabrication 

of sources, waste disposal, etc.), and 

• Periodic X-ray examinations. 
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In terms of isotopes of concern, Table 5-2 identifies the potential radiological contaminants associated 

with the specific nuclear operations at SSFL-Area IV. 

7.3.1 Process-Related External Radiation Doses at SSFL-Area IV 

The following subsections summarize the extent and limitations of information available for 

reconstructing the process-related external doses of members of the evaluated class. 

7.3.1.1 Beta and Photon Radiation Dosimetry and Source Term

Beta and photon monitoring data are available for Area IV of the SSFL in the form of personnel 

individual dosimetry records and area monitoring surveys. From the beginning of operations at Area 

IV of the SSFL, potential exposure to workers, particularly in reference to skin or extremity dose, was 

routinely considered. Exposure to beta radiation was typically accompanied by photon radiation 

exposures in SSFL-Area IV work areas. The summary of the beta and photon exposure evaluations, 

as discussed in this report and as reviewed and characterized in ORAUT-TKBS-0038-6, are included 

below. 

The beta spectrum from uranium is highly dependent on the quantity of daughter products in the 

uranium, which in turn depends on the enrichment level of the uranium. In some cases, the personnel 

dosimetry information includes extremity dosimeter results that can be considered in combination 

with the shielded dosimeter results to evaluate shallow or beta dose. The personnel beta exposures 

and doses can be reconstructed through an evaluation of the associated individual’s personnel 

monitoring records and application of dose reconstruction methods. Because dosimeter results are 

available for the radiation workers working at SSFL-Area IV, which would include the most highly 

exposed individuals, NIOSH has concluded that sufficient information exists to permit bounding the 

beta exposures for the evaluated class. 

In addition to the analysis of the beta exposures, low-energy photons also contribute to personnel 

shallow dose. Plutonium produces very few highly-penetrating gamma rays because they occur at 

such a low emission probability. Most photons from plutonium are from low-energy X-rays that are 

easily shielded. The primary source of external photon exposure is the decay of americium-241, 

which produces a 60 keV photon with 37% probability of emission per alpha disintegration. 

Americium-241 is the beta decay progeny of plutonium-241. The americium-241 in-growth in the 

plutonium increases with time, and hence increasingly contributes its higher-energy photon 

component to the total exposure. The plutonium-241 decay beta is of low energy (0.022 MeV) and is 

completely shielded by thin rubber gloves and various other materials (DOE, 2006). Based on the 

available source term information associated with these exposures, NIOSH has concluded that 

sufficient information exists to permit bounding the low-energy photon exposures for the SSFL-Area 

IV evaluated worker class.

High-energy photon exposures also resulted from SSFL-Area IV routine radiological operations with 

uranium and other high-energy photon sources (as discussed in the SSFL Technical Basis Documents 

and this evaluation report). Enriched uranium has much less in-growth of protactinium-234m, but 

uranium-235 and its decay products emit a 185.7 keV photon 57% of the time and a 143.8 keV photon 

11% of the time. These two photons dominate the measured photon energy spectra. It is appropriate 

to assume that the entire photon dose from enriched uranium is a result of exposure in the 30 - 250 

keV photon energy range. For enriched uranium, the reconstruction assumes that all photon doses 
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were a result of exposure to the 30 – 250 keV photon energy range, which provides a bounding dose 

estimate for external dose reconstruction (DOE, 2006). Because dosimeter results are available for the 

radiation workers working at SSFL-Area IV, which would include the most highly exposed 

individuals, NIOSH has concluded that sufficient information exists to permit bounding the beta 

exposures for the evaluated worker class. 

7.3.1.2 Neutron Radiation Dosimetry and Source Term

Facilities at Area IV of the SSFL with the potential for neutron radiation exposures include the 

reactors, accelerators, and fuel storage facilities. AI operated a Van de Graaff generator in Building 

4030 to bombard tritium targets with deuterons to produce neutrons. A second Van de Graaff 

generator operated at the SRE Facility to generate neutrons for neutron activation analyses of 

materials (1960 through 1964). The Water Boiler Neutron Source (WBNS), a small low-power 

research reactor, was used for reactor operator training and as a neutron source for many different 

tests from 1956 through 1980 (ORAUT-TKBS-0038-2). From the perspective of personnel neutron 

exposures, the accelerators were the major source of neutrons. 

Because of strict criticality controls at the site, no neutron-induced multiplication from plutonium 

would be expected at Area IV of SSFL. Induced fission seems to be a problem only in metal (1 kg or 

more) or in very large, high-density arrays of plutonium oxide with an additional moderator. The 

spontaneous fission neutron fields associated with plutonium at Area IV of SSFL varied based on the 

amount of shielding (or moderator material) around the plutonium. Most spontaneous fission 

neutrons produced by plutonium have energies less than 20 MeV, with the majority of energies being 

less than 5 MeV (DOE, 2006). 

Beginning with the start of reactor operations and the use of Van de Graaff accelerators, neutron doses 

were measured using NTA film. Neutron doses were recorded as whole-body dose in rem and are 

included in the records submitted to NIOSH for those claimants with the potential for neutron 

exposures. The general trend in recorded neutron dose for SSFL workers is available in the DOE 

CEDR external dosimetry analysis file for the epidemiologic study conducted by Boice, et al. (2006). 

The earliest positive neutron dose occurred in 1954, and neutron doses were measured in all years 

until 1987. NTA is known to have limitations for lower-energy neutrons (<~0.5 MeV). Because 

most, if not all, of the neutron energies that existed at Area IV of SSFL were >0.5 MeV (ORAUTTKBS-00038-6), the available NTA measurement results are expected to be a reasonable 

representation of the SSFL neutron exposures and associated doses. 

In the case that it becomes necessary to evaluate exposures to, and doses from, lower-energy neutrons 

at Area IV of SSFL, it would be possible to evaluate the neutron-to-photon ratio to account for and 

assign a neutron dose (considering that it was the AI practice at Area IV of the SSFL to assign 

beta/photon personnel dosimeters to all workers entering a radiologically controlled area; therefore, 

any significant photon radiation dose would have also been measured in tandem with the neutron 

dose). NIOSH also has access to area monitoring data that could be used to supplement and verify 

any neutron-to-photon ratios, as necessary. 

Because NTA dosimeter results are available for all applicable members (radiation workers who 

worked in areas where neutron exposures were expected) of the SSFL-Area IV worker class evaluated 

in this report, and NIOSH has knowledge of the associated source terms that result in appreciable 
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neutron exposures, NIOSH has concluded that sufficient information exists to permit bounding the 

neutron exposures for the evaluated class. 

7.3.1.3 Application of Coworker Data for External Dose Reconstruction

In cases of lost or damaged dosimeters, dose results are derived from past results of similar work, 

coworker results, or the product of instrument measurements and time spent in the radiation zone. 

These practices are typical of sites with similar circumstances. 

Coworker data may be used for cases not having complete monitoring data and may fall into one of 

several categories including: 

• The worker was unmonitored, and even by current standards did not need to be monitored (e.g., a 

non-radiological worker) 

• The worker was unmonitored, but by current standards would have been monitored 

• The worker may have been monitored, but the data are not available for dose reconstruction 

• The worker has partial information, but it is insufficient to complete a dose reconstruction 

Although a significant dosimetry data deficiency has not been identified for the evaluated worker class 

timeframe, NIOSH has proceeded to complete an external coworker model based on the available 

digitized data available to NIOSH from the study by Boice et al. (2006, May 2006). The coworker 

model is deemed useful for cases such as outlined in the paragraph above. 

7.3.2 Ambient Environmental External Radiation Doses at SSFL 

NIOSH has determined that ambient environmental external dose information is not available during 

the period evaluated in this report (sufficient data are not available until the mid-1970s). Using the 

available data, NIOSH estimated that the baseline dose above background in Area IV is 20 

mrem/year, to account for reactor operations during the period from 1955 through 1965. Facilityspecific dose for the RMDF during these years was estimated to be elevated to 80 mrem/year. During 

the years of SRE operations, a facility-specific dose of 40 mrem/year was estimated for Building 4143 

(ORAUT-TKBS-0038-4). 

The use of facility ambient external monitoring data for purposes of assessing external ambient 

environmental radiation doses was evaluated for purposes of performing EEOICPA dose 

reconstructions. However, NIOSH has concluded that the ambient environmental method (described 

above) does not support a bounding external exposure evaluation method for the 1955-1958 

timeframe, in support of radiological dose reconstructions for the proposed worker class. However, 

the available data including the external coworker model can be used to bound the external ambient 

doses for unmonitored workers during the remainder of the evaluation period (1959-1965). Therefore, 

further evaluation of the ambient environmental external doses is not included in this evaluation 

report. 
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7.3.3 SSFL-Area IV Occupational X-Ray Examinations 

AEC-funded work at SSFL-Area IV began in 1948. Clinicians prescribed chest X-ray examinations, 

and in some cases lumbar-spine X-ray examinations. These examinations caused exposure to the 

lungs and other tissues of the body. Exposure came from the primary X-ray beam and from scattered 

and leakage radiation. Radiation safety standards dating back to 1966 required “pre-exposure 

examinations” for radiation workers (Garcia, 1965). This examination was to include a chest X-ray. 

It is reasonable and claimant-favorable to assume that this practice has been in effect for radiation 

workers since the start of operations in Area IV. 

A review of randomly selected X-ray films dating back to 1956 provided insight into the occupational 

medical program (Morris, 2005). Individual medical charts do not contain X-ray films. If an old film 

is to be reviewed, the separate X-ray record must be retrieved. Three boxes of archived radiographs 

representing approximately 300 individuals were reviewed. The boxes contained an envelope for each 

individual; each envelope contained films for that person. The outside of each file storage envelope 

describes the contents; thus, it is possible to determine the number and kind of radiograph, and 

whether the radiograph represents a pre-employment examination, periodic re-examination, or a 

workup following an industrial injury. Radiographs associated with an industrial injury are often 

identified as “IND;” these are not pertinent to dose reconstruction. 

Lacking specific information in a claimant record, a reasonable and claimant-favorable assumption is 

that all employees had a pre-employment examination consisting of the following: 

• PA chest film, 

• LAT chest film, 

• AP lumbar spine film, and 

• AP lumbar spine spot film. 

7.3.4 External Dose Reconstruction 

NIOSH uses an established protocol for assessing external exposure when performing dose 

reconstructions (these protocol steps are discussed in the following subsections): 

• Photon Dose 

• Electron Dose 

• Neutron Dose 

• Unmonitored Individuals Working in Production Areas 

• Medical X-ray 
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7.3.4.1 Photon Dose

Standard external dose reconstruction practices can be employed by using the actual individual 

dosimeter results and by assigning missed dose to each non-positive dosimeter cycle (ORAUT-TKBS0038-6; ORAUT-OTIB-0004; ORAUT-OTIB-0020). 

7.3.4.2 Electron Dose

As discussed in Section 5.0, although it is not expected that beta radiation would be a significant 

exposure concern or contributor to the personnel dose for the worker class evaluated in this report, 

dosimeters used during the time period under evaluation included an open window for assessing 

shallow dose from electrons. Actual individual dosimeter results may be used to reconstruct skin dose 

when applicable (ORAUT-TKBS-0038-6; ORAUT-OTIB-0017). 

7.3.4.3 Neutron Dose

Facilities with the potential for exposure to neutron radiation included reactors, accelerators, and fuel 

storage facilities. NTA film was incorporated into the film dosimeters if there was a potential for 

exposure >100 mrem in those facilities where neutron exposures were possible (recorded in the dose 

record, under the “n” column). It is assumed that the dose recorded was the result of fast neutron 

exposure (ORAUT-TKBS-0038-6, pages 8-10). The distribution of energies and International 

Commission on Radiological Protection (ICRP) Publication 60 conversion correction factors (ICRP, 

1991; ORAUT-OTIB-0055) are listed in ORAUT-TKBS-0038-6. The correction factor for the 2- to 

14-MeV energy group was calculated from data in ORAUT-TKBS-0014-6, based on the similarities 

in the badges used at the two facilities. 

7.3.4.4 Medical X-ray

The records provided by DOE typically include adequate information to define the date, type, and 

count of X-ray examinations that were administered to each employee as a condition of employment. 

Assumptions regarding radiographic exposure frequency are used only for screening or when specific 

claimant records are not available. 

In accordance with ORAUT-OTIB-0006 and ORAUT-PROC-0061, NIOSH has data and information 

to support bounding the medical X-ray dose. 

7.3.5 External Dose Reconstruction Feasibility Conclusion 

NIOSH has established that it has access to sufficient information to either: (1) estimate the maximum 

external radiation dose for every type of cancer for which radiation doses are reconstructed that could 

have been incurred under plausible circumstances by any member of the class; or (2) estimate the 

external radiation doses to members of the class more precisely than a maximum dose estimate. 

7.4 Evaluation of Petition Basis for SEC-00093 

The following subsections evaluate the assertions made on behalf of petition SEC-00093 for SSFLArea IV. 
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7.4.1 Evaluation of Specific Petitioner Statements in SEC-00093 

This subsection presents general concerns made by workers on behalf of petition SEC-00093. The 

italicized statements are from the petition; the comments that follow are from NIOSH. 

7.4.1.1 SRE Incident and Release of Core Gases 

The petitioner indicated that based on information about the SRE partial meltdown of fuel rods, it was 

the petitioner’s intent to maintain this incident as part of their petition basis that monitoring records 

do not exist. On the Continuation page of Form B—under See Enclosures— the petitioner listed 

“release of core gases.” 

In addition to reference documents located in the SRDB, NIOSH has reviewed several documents that 

specifically evaluate the environmental releases resulting from this incident (Christian, 2005; 

Lochbaum, 2006; Hart, 1962, Rutherford, 2005). Although there is not a consensus on the exact 

amount of gaseous radioactive materials released to the environment following the incident, all 

pertinent scientific reports conclude that a significant amount of fission products was released from 

the fuel into the primary sodium coolant, with a percentage of that inventory being released to the 

reactor’s cover gas system, and subsequently released into the atmosphere through reactor building 

vent systems and from gaseous storage tanks. The type and range of releases to the environment 

following the fuel damage run from less than 1 Ci of iodine-131 (Christan, 2005) to a conservative 

upper-bound estimate of ~ 3000 Ci of iodine-131 (Lochbaum, 2006). Based on documented stack 

releases during the incident, AI concluded that since no iodine-131 was detected in the cover gas, only 

about 28 Ci of the noble gases krypton-85 and xenon-133 were released from the stacks to the 

environment (Rutherford, 2005). 

Based on its review of claims in NOCTS, NIOSH concludes that personnel monitoring does exist for 

members of the evaluated class (both internal monitoring and external monitoring) during the 

timeframe of the SRE event. In addition to the personnel monitoring, some air monitoring 

measurements from the reactor area and stack monitoring also exist. The availability of actual 

bioassay results (and associated program data and documentation that support the ability to interpret 

the internal data) post-1958 for affected and/or associated members of the evaluated worker class 

supports NIOSH’s ability to reconstruct dose for those members of the evaluated class with sufficient 

accuracy. Because these data would include the worst-case exposure scenarios for the evaluated 

worker class, NIOSH contends that the availability of these data also supports its ability to establish a 

bounding exposure scenario for the entire evaluated worker class after 1958. At this point NIOSH has 

completed the development of an external coworker model to assess potential doses to workers who 

were not monitored, but should have been monitored based on their employment location or potential 

unplanned involvement in the incident response. NIOSH is currently in the process of developing an 

internal coworker model for the same purpose. 

7.4.1.2 Radiation Badges 

The petitioner discussed a Tiger Team report indicating that it detailed “inadequate radiation 

badges.” 
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The Tiger Team Report cited by the petitioner does not state that the dosimeters were inadequate, but 

that they were not Department of Energy Laboratory Accreditation Program (DOELAP)-accredited, 

and it specifically focuses on the D&D period versus the period being evaluated in this report. As it 

relates to the program being discussed in the Tiger Team report, it was common practice (and is noted 

as an option in Section 7 of DOELAP Administrative Standard DOE-STD-1111-98) for smaller 

programs to be exempted from DOELAP accreditation, contingent upon using a National Voluntary 

Laboratory Accreditation Program (NVLAP)-accredited commercial service (DOE, 2006). However, 

this is pertinent only beginning in 1986 when the DOELAP requirements were implemented. 

Therefore, as discussed previously, the period cited in the petition is after the period of the evaluation 

and does not impact this Evaluation Report. 

7.4.1.3 Tritium Plumes 

On the Continuation Page of Form B—under See Enclosures—the petitioner listed “tritium plumes.” 

Available data indicate that onsite water supply wells were the primary water source from 1949 to 

1964. No detectable (>1000 pCi/l) tritium has ever been found in any of the water supply wells. The 

primary supply wells (WS-5, WS-6, WS-12, and WS-13) were in Areas I, II, and III at SSFL. One 

well (WS-7) was in Area IV and was a minor source compared to the other wells. Recent sampling 

indicates that the source of the tritium is in the vicinity of Building 4010. None of the sampling wells 

between the source and the historical water supply wells has shown detectable levels of tritium (>1000 

pCi/l) (Haley & Aldrich, 2007). Based on the available data, NIOSH can bound occupational 

exposure to tritium in the groundwater by assuming that the concentrations in the shallow sampling 

well RD-34A are related to concentrations in the drinking water wells in the 1950s and 1960s. It 

should be noted that sampling well RD-34A has never been a source of drinking water for the site and 

is down-gradient from the source of the tritium (Building 4010). Since 1991, the mean concentration 

in this well has been 2,940 pCi/l. For the purpose of establishing a bounding approach to 

reconstructing dose from this exposure source to the evaluated worker class evaluated in this report, it 

can be assumed that the tritium instantaneously built up in the soil and was transported to the 

groundwater. This would result in a concentration of 30,000 pCi/l in the 1950s and 1960s (accounting 

for radioactive decay). Exposures would be considered incidental after the point that bottled drinking 

water was delivered to the site (1962 and later). 

Starting in 1964, offsite water was imported into the pipelines (public water supply), and by 1965 the 

majority of the water supplying the restrooms, kitchens, etc. was from an offsite source. All supply 

wells were sampled for tritium in the 1980s and 1990s, and none was detected. Around 2000, the 

imported water that supplied the restrooms, etc. was completely separated from the supply wells 

onsite (Haley & Aldrich, 2007; Rutherford, 2007). 

7.4.1.4 Uranium Fires 

On the Continuation Page of Form B—under See Enclosures—the petitioner listed “uranium fires.” 

Documentation supporting the petitioner’s claim cites two incidents of a sodium explosion while 

cleaning a valve, and a uranium carbide oxidation incident. NIOSH has assumed that the petitioner 

has provided these as examples of incidents without monitoring data. The documents submitted 

contain data indicating that radiological surveys of Building 064 Vault and the CERF (Building 163) 

were performed following each of these events. Monitoring consisted of survey meters and smears 
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that indicated beta contamination levels of 726 d/m – 57,762 d/m-100cm2 for Building 163 and 5 x 

103

 d/m-100 cm2 for Building 064. Building 163 had air monitoring data with concentrations of 2 x 

10-9 μCi/cm3

 following the events (Name One, September 6, 2007). Based on the available 

information for these incidents, coupled with the available personnel monitoring data, NIOSH 

contends that it can bound the dose associated with these exposures for the evaluated worker class 

evaluated in this report. A review of 38 additional uranium fire incident reports indicate that the areas 

were checked after the incidents for contamination, and those individuals involved in the incidents 

went through decontamination and medical checks with no injuries reported. 

7.4.1.5 Air Monitoring

On the Continuation Page of Form B—under See Enclosures— the petitioner listed “no air 

monitoring.” 

Based on NIOSH’s review of the information available in the SRDB, the Site Profile, and the 

personnel monitoring data in NOCTS, NIOSH concurs that there is an apparent lack of area air 

monitoring for the period at SSFL-Area IV prior to 1958, which does impact the feasibility of 

estimating internal radiation doses with sufficient accuracy for the proposed worker class during that 

time period. 

After 1958, NIOSH will base its dose reconstructions primarily on bioassay data. The availability of 

actual bioassay results (and associated program data and documentation that supports the ability to 

interpret the internal data) for members of the evaluated worker class supports NIOSH’s ability to 

reconstruct dose for those members of the evaluated class with sufficient accuracy. Because these 

data would include the worst-case exposure scenarios for the evaluated worker class, NIOSH contends 

that the availability of this data also supports the ability to establish a bounding exposure scenario, 

regardless of the availability of air sample data, for the evaluated worker class after 1958, when 

bioassay data became routinely available. 

7.4.1.6 The Sodium Burn Pit

As identified in Item F.1 of the SEC-00093 Form B, the petitioner discussed the Sodium Burn Pit and 

indicated that no records were kept. 

The Sodium Burn Pit was used to clean sodium from equipment parts before disposal and was not 

intended to be used for disposal of radioactive material. The site profile does not evaluate the 

potential emissions from these operations (due to contamination of the sodium with fission products). 

However, it does state that “several small scrap test components containing radioactivity were found 

and removed” (ORAUT-TKBS-0038-2). 

The Interim Final RCRA Facility Assessment Report for Rockwell International Corporation states 

that the Sodium Burn Pit had soil contamination, but it was unknown if radioactive waste had been 

burned there. In October 1978, three contaminated sodium barrels were found in the sodium burn pit. 

The barrels had a maximum exposure rate of 1 mrad/hour. The barrels were removed and taken to the 

RMHF for disposal (Sapere Consulting, 2005). After the discovery of contaminated items at the site, 

periodic radiation surveys and soil samples were performed from 1978 through 1983. Based on 

process history, the contaminants of concern are cesium-137, strontium, and thorium. Results 
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indicated low levels (56 pCi/g maximum) of radioactive contamination (principally cesium-137), 

primarily in the lower pond. Contamination was not identified in areas outside the ponds. 

The Sodium Burn Pit was an outdoor area, neither continuously occupied nor continuously used. 

Because the pit was not an operations area, exposures for individuals that may have intermittently 

occupied the Sodium Burn Pit would have been lower than exposures for individuals who performed 

operations work with the source materials that were delivered to this location. In addition, significant 

radiological exposures resulting from worker activities in the vicinity of the Sodium Burn Pit are 

unlikely because of the controls in place at this location. For example, workers were required to 

maintain a safe distance from the pits (including lined and unlined pits/ponds) because of the 

potentially violent reactions that could occur in the case of sodium or potassium making contact with 

water. 

The availability of actual bioassay results after 1958 (and associated program data and documentation 

that support the ability to interpret the internal data) supports NIOSH’s ability to reconstruct dose for 

those workers who might have been affected by or associated with burn pit operations. Because the 

data would include the most highly exposed individuals , NIOSH contends that the availability of 

these data also supports the ability to establish a bounding exposure scenario for any exposures to 

personnel at the burn pit, or for any member of the worker class evaluated in this report and who 

worked after the 1955-1958 time period. 

7.5 Other Issues Relevant to the Petition Identified During the Evaluation 

During the feasibility evaluation for SEC-00093, a number of issues were identified that required 

further analysis and resolution. The issues and their current status are as follows: 

• ISSUE: The identification of workers with blank radiation exposure record sheets in their file (a 

sheet with no entries). 

APPROACH: In an interview with a current Radiation Safety Officer at the SSFL, the interviewee 

stated that each AI worker (both radiation and non-radiation workers) had a blank record sheet in 

his/her file called a “blue card.” If they were assigned to a controlled area, they were required to 

have a film badge and any exposure was entered into their file (Personal Communication with 

Health Physicist, November 20, 2007). This information was corroborated in random individual 

record reviews performed by NIOSH as well as in documented reviews of personnel records 

(Boice, 2006; Boice, May 2006); therefore, NIOSH has concluded that individuals that had no 

record (or blue card) are considered non-radiological workers at Area IV of the SSFL. 

Based on information included in documented reviews (Boice, 2006; Boice, May 2006), during 

the early years at the site all personnel were issued a radiation dose folder/record; however, only 

radiological workers (who were issued dosimeters) received recorded dose information in those 

records. Based on this information, NIOSH has concluded that individuals with blank radiological 

folders/records are also non-radiological workers (at Area IV of SSFL), and individuals with 

recorded information in their radiological files are considered SSFL-Area IV radiological workers. 

As previously discussed, the availability of personnel records for monitored individuals supports 

NIOSH’s ability to reconstruct dose with sufficient accuracy for the evaluated worker class 

members for the period after 1958. Because the available data also include a representation of the 

maximum potential exposures (a bounding exposure scenario) for the evaluated worker class, 
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NIOSH contends that this supports its ability to bound the associated dose for all members of the 

evaluated worker class after 1958, including any unmonitored individual or any individual with 

potentially missing dosimetry data (this approach will be handled on an individual case basis). 

• ISSUE: When firemen from other sites were involved in fires or events at SSFL Area IV, how was 

it recorded? 

APPROACH: Based on interviews and reviews of dosimetry information for this evaluation, 

NIOSH has determined that firemen typically wore film badges when working in areas with the 

potential for radiological exposures. In one case, an individual that worked as a fireman did not 

have monitoring records in his file. During interviews, interviewees indicated that all firemen 

wore film badges regardless of where they were assigned for a shift. One interviewee stated that 

“During the 30 years I was there (1962 through 1993), there were no radiation fires” (Personal 

Communication with Fireman). Based on the information from interviews, it is clear that each 

individual claim will have to be reviewed and processed on a case-by-case basis; it is possible that 

the fireman without monitoring records was not expected to receive any exposures while on the 

job and was not required to enter radiological areas during his employment. However, this does 

not prevent NIOSH from establishing a bounding exposure scenario for the evaluated worker 

class. As previously discussed and stated, the availability of personnel records for monitored 

individuals supports NIOSH’s ability to reconstruct dose with sufficient accuracy for workers after 

1958. Because the available data also include a representation of the maximum potential 

exposures (a bounding exposure scenario), NIOSH contends that this supports its ability to bound 

the associated dose for all members of the evaluated worker class after 1958, including dose 

associated with the exposure scenarios presented by/for these firemen. 

7.6 Summary of Feasibility Findings for Petition SEC-00093 

This report evaluates the feasibility for completing dose reconstructions for employees at SSFL-Area 

IV from January 1955 through December 1965. Based on the limited pre-1959 internal monitoring 

data and associated program or source term information, NIOSH has concluded that there is 

insufficient information to support establishing a bounding internal exposure scenario for the 

proposed worker class that worked in Area IV of the SSFL from 1955 through 1958. NIOSH has 

identified post-1958 radiological internal monitoring program data and the original monitoring data 

for individuals working in the highest-exposure areas at Area IV of the SSFL. Therefore, NIOSH has 

concluded that there are sufficient data for it to bound the dose (reconstruct dose with sufficient 

accuracy) for the class evaluated for the period from 1959 through 1965. 

Table 7-1 summarizes the results of the feasibility findings at SSFL-Area IV for each exposure source 

during the time period of January 1955 through December 1965. 

Table 7-1: Summary of Feasibility Findings for SEC-00093 

January 1955 through December 1965 

Source of Exposure Reconstruction Feasible Reconstruction Not Feasible 

Internal1 X 

(January 1, 1959 – December 31, 1965) 

X 

(January 1, 1955 – December 31, 1958) 

 - All Radionuclides X X 
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Table 7-1: Summary of Feasibility Findings for SEC-00093 

January 1955 through December 1965 

(January 1, 1959 – December 31, 1965) (January 1, 1955 – December 31, 1958) 

External X 

 - Beta, Photon, and Neutron X 

 - Occupational Medical X-ray X 

Notes: 

1

 Internal includes an evaluation of urinalysis (in vitro) and airborne dust data. 

8.0 Evaluation of Health Endangerment for Petition SEC-00093 

The health endangerment determination for the class of employees covered by this evaluation report is 

governed by both EEOICPA and 42 C.F.R. § 83.13(c)(3). Under these requirements, if it is not 

feasible to estimate with sufficient accuracy radiation doses for members of the class, NIOSH must 

also determine that there is a reasonable likelihood that such radiation doses may have endangered the 

health of members of the class. Section 83.13 requires NIOSH to assume that any duration of 

unprotected exposure may have endangered the health of members of a class when it has been 

established that the class may have been exposed to radiation during a discrete incident likely to have 

involved levels of exposure similarly high to those occurring during nuclear criticality incidents. If 

the occurrence of such an exceptionally high-level exposure has not been established, then NIOSH is 

required to specify that health was endangered for those workers who were employed for a number of 

work days aggregating at least 250 work days within the parameters established for the class or in 

combination with work days within the parameters established for one or more other classes of 

employees in the SEC. 

Per EEOICPA and 42 C.F.R. § 83.13(c)(3), a health endangerment determination is required because 

NIOSH has determined that it lacks sufficient internal monitoring data for the period from January 1, 

1955 through December 31, 1958, which prevents evaluating internal dose with sufficient accuracy 

for the members of the proposed class during this period. 

NIOSH has determined that members of the class were not exposed to radiation during a discrete 

incident likely to have involved levels of exposure similarly high to those occurring during nuclear 

criticality incidents. However, evidence indicates that some workers in the proposed class may have 

accumulated substantial chronic exposures through episodic intakes of radionuclides, combined with 

external exposures to gamma, beta, and neutron radiation. Consequently, NIOSH is specifying that 

health was endangered for those workers covered by this evaluation who were employed for a number 

of work days aggregating at least 250 work days within the parameters established for this class or in 

combination with work days within the parameters established for one or more other classes of 

employees in the SEC (excluding aggregate work day requirements).

9.0 NIOSH-Proposed Class for Petition SEC-00093 

Based on its research, NIOSH reduced the petitioner-requested class to define a single class of 

employees for which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSHproposed class includes all employees of the Department of Energy (DOE), its predecessor agencies, 

and DOE contractors and subcontractors who worked in any area of Area IV of the Santa Susana Field 
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Laboratory for a number of work days aggregating at least 250 work days from January 1, 1955 

through December 31, 1958, or in combination with work days within the parameters established for 

one or more other classes of employees in the SEC (excluding aggregate work day requirements). 

NIOSH has carefully reviewed all material sent in by the petitioner, including the specific assertions 

stated in the petition, and has responded herein (see Section 7.4). NIOSH has also reviewed available 

technical resources and many other references, including the Site Research Database (SRDB), for 

information relevant to SEC-00093. In addition, NIOSH reviewed its NOCTS dose reconstruction 

database to identify EEOICPA-related dose reconstructions that might provide information relevant to 

this petition evaluation. 

These actions are based on existing, approved NIOSH processes used in dose reconstruction for 

claims under EEOICPA. NIOSH’s guiding principle in conducting these dose reconstructions is to 

ensure that the assumptions used are fair, consistent, and well-grounded in the best available science. 

Simultaneously, uncertainties in the science and data must be handled to the advantage, rather than to 

the detriment, of petitioners. When adequate personal dose monitoring information is not available, or 

is very limited, NIOSH may use the highest reasonably possible radiation dose, based on reliable 

science, documented experience, and relevant data to determine the feasibility of reconstructing the 

dose of an SEC petition class. NIOSH contends that it has complied with these standards of 

performance in determining that it would not be feasible to reconstruct the internal dose for the class 

proposed in this petition. 
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10.0 References 

42 C.F.R. pt. 81, Guidelines for Determining the Probability of Causation Under the Energy 

Employees Occupational Illness Compensation Program Act of 2000; Final Rule, Federal 

Register/Vol. 67, No. 85/Thursday, p 22,296; May 2, 2002; SRDB Ref ID: 19391 

42 C.F.R. pt. 82, Methods for Radiation Dose Reconstruction Under the Energy Employees 

Occupational Illness Compensation Program Act of 2000; Final Rule; May 2, 2002; SRDB Ref ID: 

19392 

42 C.F.R. pt. 83, Procedures for Designating Classes of Employees as Members of the Special 

Exposure Cohort Under the Energy Employees Occupational Illness Compensation Program Act of 

2000; Final Rule; May 28, 2004; SRDB Ref ID: 22001 

42 U.S.C. §§ 7384-7385 [EEOICPA], Energy Employees Occupational Illness Compensation 

Program Act of 2000, as amended 

H. R. 2242, copy of H. R. 2243; 110th Congress, 1st Session; May 9, 2007; OSA Ref ID: 103253

OCAS-IG-001, External Dose Reconstruction Implementation Guideline, Rev. 01; Office of 

Compensation Analysis and Support (OCAS); Cincinnati, Ohio; August 2002; SRDB Ref ID: 22401 

OCAS-PR-004, Internal Procedures for the Evaluation of Special Exposure Cohort Petitions, Rev. 0; 

National Institute for Occupational Safety and Health (NIOSH); Cincinnati, Ohio; September 23, 

2004; SRDB Ref ID: 32022 

ORAUT-OTIB-0004; Estimating the Maximum Plausible Dose to Workers at Atomic Weapons 

Employer Facilities, Rev. 03 PC-2; Oak Ridge Associated Universities (ORAU); Oak Ridge, 

Tennessee; December 6, 2006; SRDB Ref ID: 29949 

ORAUT-OTIB-0006; Dose Reconstruction from Occupationally Related Diagnostic X-Ray 

Procedures; Rev. 03; Oak Ridge Associated Universities (ORAU); Oak Ridge, Tennessee; August 2, 

2005; SRDB Ref ID: 19422

ORAUT-OTIB-0010; A Standard Complex-Wide Methodology for Overestimating External Doses 

Measured with Film Badge Dosimeters; Rev. 01; Oak Ridge Associated Universities (ORAU); Oak 

Ridge, Tennessee; June 5, 2006; SRDB Ref ID: 29953 

ORAUT-OTIB-0017; Interpretation of Dosimetry Data for Assignment of Shallow Dose, Rev. 00; 

Oak Ridge Associated Universities (ORAU); Oak Ridge, Tennessee; January 19, 2005; SRDB Ref ID: 

19435

ORAUT-OTIB-0018; Internal Dose Overestimates for Facilities with Air Sampling Programs; Rev. 

01; Oak Ridge Associated Universities (ORAU); Oak Ridge, Tennessee; August 9, 2005; SRDB Ref 

ID: 19436 
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ORAUT-OTIB-0020; Use of Coworker Dosimetry Data for External Dose Assignment, Rev. 00; Oak 

Ridge Associated Universities (ORAU); Oak Ridge, Tennessee; December 29, 2004; SRDB Ref ID: 

19441

ORAUT-OTIB-0033; Application of Internal Doses Based on Claimant-Favorable Assumptions for 

Processing as Best Estimates; Rev. 00; Oak Ridge Associated Universities (ORAU); Oak Ridge, 

Tennessee; April 20, 2005; SRDB Ref ID: 19457 

ORAUT-OTIB-0054, Fission and Activation Product Assignment for Internal Dose-Related Gross 

Beta and Gross Gamma Analyses, Rev. 00, Oak Ridge Associated Universities (ORAU); Oak Ridge, 

Tennessee; May 11, 2007; SRDB Ref ID: 31780 

ORAUT-OTIB-0055, Technical Basis for Conversion from NCRP Report 38 Neutron Quality Factors 

to ICRP Publication 60 Radiation Weighting Factors for Respective IREP Input Neutron Energy 

Ranges, Rev. 00, Oak Ridge Associated Universities (ORAU); Oak Ridge, Tennessee; June 5, 2006; 

SRDB Ref ID: 29980 

ORAUT-PROC-0060; Occupational Onsite Ambient Dose Reconstruction for DOE Sites, Rev. 01; 

Oak Ridge Associated Universities (ORAU); Oak Ridge, Tennessee; June 28, 2006; SRDB Ref ID: 

29986 

ORAUT-PROC-0061; Occupational X-Ray Dose Reconstruction for DOE Sites, Rev. 01; Oak Ridge 

Associated Universities (ORAU); Oak Ridge, Tennessee; July 21, 2006; SRDB Ref ID: 29987 

ORAUT-TKBS-0006-6, Technical Basis Document for the Hanford Site—Occupational External 

Dosimetry, Rev. 00; Oak Ridge Associated Universities (ORAU); Oak Ridge, Tennessee; October 20, 

2003; SRDB Ref ID: 19497 

ORAUT-TKBS-0014-6, Technical Basis Document for the Y-12 National Security Complex—

Occupational External Dosimetry, Rev. 00 PC-3; Oak Ridge Associated Universities (ORAU); Oak 

Ridge, Tennessee; May 11, 2006; SRDB Ref ID: 30048 

ORAUT-TKBS-0038-1; Atomics International – Introduction, Rev. 01; Oak Ridge Associated 

Universities (ORAU); Oak Ridge, Tennessee; August 30, 2006; SRDB Ref ID: 30080 

ORAUT-TKBS-0038-2; Energy Technology Engineering Center – Site Description, Rev. 00; Oak 

Ridge Associated Universities (ORAU); Oak Ridge, Tennessee; February 2, 2006; SRDB Ref ID: 

22140 

ORAUT-TKBS-0038-3; Area IV of the Santa Susana Field Laboratory, the Canoga Avenue Facility 

(Vanowen Building), the Downey Facility, and the De Soto Avenue Facility (sometimes referred to as 

Energy Technology Engineering Center)– Occupational Medical Dose, ORAUT-TKBS-0038-3; Rev. 

02; October 31, 2008; SRDB Ref ID: 53184 

ORAUT-TKBS-0038-4; Area IV of the Santa Susana Field Laboratory, the Canoga Avenue Facility 

(Vanowen Building), the Downey Facility, and the De Soto Avenue Facility (sometimes referred to as 

Energy Technology Engineering Center [ETEC] or Atomics International) – Occupational 
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Environmental Dose, Rev. 01; Oak Ridge Associated Universities (ORAU); Oak Ridge, Tennessee; 

March 8, 2007; SRDB Ref ID: 30622 

ORAUT-TKBS-0038-5; Energy Technology Engineering Center – Occupational Internal Dose, Rev. 

00; Oak Ridge Associated Universities (ORAU); Oak Ridge, Tennessee; February 22, 2006; SRDB 

Ref ID: 22287 

ORAUT-TKBS-0038-6; Atomics International – Occupational External Dose, Rev. 01; Oak Ridge 

Associated Universities (ORAU); Oak Ridge, Tennessee; November 16, 2006; SRDB Ref ID: 30082 

A Briefing for Elected Officials on the Health Risks to the Public from SSFL, DVD submitted by 

petitioner, dated April 5, 2007

Addeo, 1960, Monthly Report—March, 1960; Chemical Disposal—Activities at Pits, Inspections, 

Burning, Convoying, Etc.; Inter-office letter from E. G. Addeo to R. J. Madden; E. G. Addeo; April 4, 

1960; OSA Ref ID: 103253 

Alexander, 1959, Radiochemical Analysis for Radioisotopes in Biological Specimens; R. E. 

Alexander; March 25, 1959; SRDB Ref ID: 19108 

Badger, 1965-1987, Early External Dosimetry Procedures at Area IV of the SSFL; F. H. Badger; 

1965-1987; SRDB Ref ID: 19062. 

Boeing, 2006, Comments of the Boeing Company on the Reports of the Santa Susana Field 

Laboratory Advisory Panel; The Boeing Company; November 3, 2006; SRDB Ref ID: 37245 

Boice, 1990, Studies of Atomic Bomb Survivors—Understanding Radiation Effects, Editorial in the 

Journal of the American Medical Association, John D. Boice; August 1, 1990; OSA Ref ID: 103253 

Boice, 2006, Mortality among Radiation Workers at Rocketdyne (Atomics International), 1948-1999; 

John D. Boice, Sarah S. Cohen, Michael T. Mumma, Elizabeth Dupree Ellis, Keith F. Eckerman, 

Richard W. Leggett, Bruce B. Boecker, A. Betrand Brill, and Brian E. Henderson; 2006; SRDB Ref 

ID: 39396 

Boice, May 2006, A Comprehensive Dose Reconstruction Methodology for Former 

Rocketdyne/Atomics International Radiation Workers; John D. Boice, Richard W. Leggett, Elizabeth 

Dupree Ellis, Phillip W. Wallace, Michael Mumma, Sarah S. Cohen, A. Bertrand Brill, Bandana 

Chadda, Bruce B. Boecker, R. Craig Yoder, and Keith F. Eckerman; Health Physics, 90 (5): 409-430; 

May 2006; SRDB Ref ID: 37539

Brownley, 2007, Santa Susana Field Lab (SSFL) Closure Plan, letter from Julia Brownley; Julia 

Brownley, Assembly Member of the 41st District; May 4, 2007; OSA Ref ID: 103253 

Cavanaugh, 2004, High-Level Radioactivity Found in Groundwater at Rocketdyne Lab; Kerry 

Cavanaugh; May 19, 2004; OSA Ref ID: 103253

Cavanaugh, 2005, Higher Tritium Levels Found in Water at Field Lab; Kerry Cavanaugh; 2005; OSA 

Ref ID: 103253
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Cavanaugh, 2007; Under Pressure, DOE Halts Field Lab Cleanup for 45 Days, newspaper article 

submitted with petition; Kerry Cavanaugh; May 26, 2007; OSA Ref ID: 103253 

Christian, 2005, Chemical Behavior of Iodine-131 During SRE Fuel Element Damage in July 1959, 

response to plaintiff’s expert witness Arjun Makhijani; Jerry D. Christian; May 26, 2005; SRDB Ref 

ID: 35752 

Clow, 1966, Neutron Survey of SRE Hot Cell Containing a 14 MeV Neutron Generator, 

correspondence to E. T. Bramlitt; H. E. Clow; February 3, 1966; SRDB Ref ID: 19061 

Curwood, 2006, Living on Earth Radio Program, interview that aired the week of January 20, 2006— 

containing comments by Arjun Makhijani and [Name One Redacted]; Steve Curwood; March 6, 2006; 

OSA Ref ID: 103253

Dempsey, 1989, Site Visit to Santa Susana Field Laboratory Operated by Rockwell/Rocketdyne, site 

visit memo; Gregg D. Dempsey; July 28, 1989; OSA Ref ID: 103253 

DHS of California, unknown date, Rocketdyne Worker Health Study Summary; California Department 

of Health Services, Occupational Health Branch; date not identified; OSA Ref ID: 103253

DHS of California, 1990, Letter Regarding Request to Evaluate the Need for an Epidemiological 

Study of Workers and Community Residents Potentially Impacted by the Operations at Rockwell 

International Corporation’s Santa Susana Field Laboratories, letter to Dr. Robert Goldsmith from the 

Department of Health Services; Department of Health Services (DHS) of California; October 5, 1990; 

OSA Ref ID: 103253 

DOE, 1986, Department of Energy Standard for the Performance Testing of Personnel Dosimetry 

Systems, DOE/EH-0027; Department of Energy (DOE); December 1986; SRDB Ref ID: 12294

DOE, 1991, Tiger Team Assessment—Energy Technology Engineering Center, DOE/EH—0175; U.S. 

Department of Energy (DOE); April 1991; OSA Ref ID: 103714 

DOE, 2006, DOE Standard: Guide of Good Practices for Occupational Radiological Protection in 

Plutonium Facilities, DOE-STD-1128-98; U.S. Department of Energy (DOE); December 2006; 

SRDB Ref ID: 34061 

DOE, 2007, ETEC Closure Project; U.S. Department of Energy (DOE); last accessed on December 7, 

2007 at http://www.etec.energy.gov/history/major-operations/support-ops/fuel-storage.html 

Edelmann, 1959, Telegram to M. E. Piedimonte; A. Edelmann; SRDB Ref ID: 19056, page 14 

EPA, 1999, The U.S. EPA Announces Results of Rocketdyne’s Off-Site Sampling Program for the 

Santa Susana Field Laboratory, EPA update; U. S. Environmental Protection Agency (EPA); July 

1999; OSA Ref ID: 103253

EPA, 2003, EPA Concludes Superfund Evaluation of ETEC Area IV, Environmental Protection 

Agency (EPA); December 2003; OSA Ref ID: 103253 
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Franke, unknown date; Curriculum Vitae for Bernd Franke; OSA Ref ID: 103253

Franke, 2004, Declaration of Bernd Franke, expert testimony in legal case; February 12, 2004; OSA 

Ref ID: 103253 

Garcia, April 5, 1963, Handling of Routine Film Badges; R. R. Garcia and Carpenter; April 5, 1963; 

SRDB Ref ID: 19182 

Garcia, July 1, 1963, Modification of the Personnel Monitoring Program Rolls, PM-1; R. R. Garcia; 

July 1, 1963; SRDB Ref ID: 19181 

Garcia, July 15, 1963, Dose Determination, PM-14; R. R. Garcia and Carpenter; July 15, 1963; SRDB 

Ref ID: 19687 

Garcia, 1965, Radiation Engineering Analysis, Radiation Safety Standard Analysis—Medical 

Examinations for Radiation Workers, revised in 1970 by F. M. Staszeski; R. R. Garcia and R. E. 

Alexander; December 8, 1965; SRDB Ref ID: 19052 

Haley & Aldrich, 2007, Analysis of Potential for Tritium in Water Supply at SSFL, 19 slide 

PowerPoint presentation; Haley & Aldrich; 2007; SRDB Ref ID: Currently not assigned 

Hart, 1962, Distribution of Fission Product Contamination in the SRE, NAA-SR-6890; R. S. Hart; 

March 1, 1962; SRDB Ref ID: 39770 

Hart, 1980, Annual Review of Radiological Controls-1979; R. S. Hart; September 15, 1980; SRDB 

Ref ID: 19133 

Haskell, 1991, Statement of the Potential Health Risks Identified in the Santa Susana Field 

Laboratory Exposure Assessment Study; Charles M. Haskell; February 4, 1991; OSA Ref ID: 103253

Hostetler, 2006, Statement of the Honorable John N. Hostetler, Chairman of the Subcommittee on 

Immigration, Border Security and Claims for the December 5, 2006 Oversight Hearing on the Energy 

Employees Occupational Illness Compensation Program Act—Are We Fulfilling the Promise We 

Made to These Veterans of the Cold War When We Created the Program?; December 5, 2006; OSA 

Ref ID: 103253 

ICRP, 1975, Reference Man: Anatomical, Physiological, and Metabolic Characteristics, ICRP 

Publication 23; International Commission on Radiological Protection (ICRP); 1975; SRDB Ref ID: 

22735 

ICRP, 1991, 1990 Recommendations of the International Commission on Radiological Protection, 

Publication 60; International Commission on Radiological Protection (ICRP); 1991; SRDB Ref ID: 

11517 

Kellehar, 1966, Bioassay Program General Information, letter to R. E. Alexander; R. G. Kellehar; 

July 19, 1966; SRDB Ref ID: 19110 
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Kopenhaver, 1986, Functional Appraisal of the Radiation Safety Program at Rocketdyne; Ralph 

Kopenhaver; 1986; SRDB Ref ID: 17004 

Lang, 1960, Radiological Safety at Atomics International—A Division of North American Aviation 

Inc.; J. C. Lang; July 13, 1960; SRDB Ref ID: 19144 

Lee, 1963; Quotation on Urinalysis Services; letter to Atomics International, BioScience Laboratories; 

Norman D. Lee; July 30, 1963; SRDB Ref ID: 19673, pages 2-6. 

Lochbaum, 2006, An Assessment of Potential Pathways for Release of Gaseous Radioactivity 

Following Fuel Damage During Run 14 at the Sodium Reactor Experiment; David A. Lochbaum; 

October 5, 2006; SRDB Ref ID: 37247 

Makhijani, 2004, Declaration of Arjun Makhijani, expert testimony in legal case; February 12, 2004; 

OSA Ref ID: 103253 

Morris, 2005, ETEC Occupational Medical Exposures: Interview with Dr. Craig Conlon, documented 

Communication with Dr. Craig Conlon; Robert L. Morris; December 9, 2005; SRDB Ref ID: 20534 

Name One, March 2006; Energy Employees Occupational Illness Compensation Program Act of 

2000, letter from [Name One Redacted] to Congressman Elton Gallegly; [Name One Redacted]; 

March 1, 2006; OSA Ref ID: 103253 

Name One, June 2006, Dose Construction [sic] for [Name Two Redacted], letter from [Name One 

Redacted] to Mr. Larry Elliott; [Name One Redacted]; June 4, 2006; OSA Ref ID: 103253 

Name One, June 2007, Letter to Substantiate SEC Petition Application for the Santa Susana Field 

Laboratory; [Name One Redacted]; June 30, 2007; OSA Ref ID: 103255

Name One, August 2007, Email from Petitioner, sent to OCAS requesting document be added to 

petition; [Name One Redacted]; August 27, 2007; OSA Ref ID: 103666

Name One, September 6, 2007, Letter Regarding SSFL-Area IV Dose Reconstruction Case, includes 

dose reconstruction with attached DR Report and handwritten statements/notes; [Name One 

Redacted]; September 6, 2007; OSA Ref ID: 103702 

Name One, September 20, 2007, Detailed Outline of Materials in Support of SEC-00093, Email to 

OCAS; [Name One Redacted]; September 20, 2007; OSA Ref ID: 103808 

Name Three, unknown date, Statement from [Name Three Redacted], regarding work experiences at 

Rocketdyne; date not identified; OSA Ref ID: 103253 

O’Brien, 1959; Quote for Urinalysis for Uranium by Fluorometric Determination, letter to Atomics 

International; Robert E. O’Brien; November 10, 1959; SRDB Ref ID: 19056. 

Oldenkamp, 1989, Nuclear Operations at Rockwell’s Santa Susana Field Laboratory—A Factual 

Perspective; R. D. Oldenkamp and J. C. Mills; December 20, 1989; SRDB Ref ID: 20546 
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Owens, 1964, Incident Report Regarding Oxidation Reaction of Uranium; D. E. Owens; October 27, 

1964; OSA Ref ID: 103700 

Pardue, 1944, Photographic Film as a Pocket Radiation Dosimeter, CH-1553-A-2223; L. A. Pardue, 

N. Goldstein, and E. O. Wollan; April 8, 1944; SRDB Ref ID: 8599 

Personal Communication with Fireman; Telephone Interview by ORAU Team; November 15, 2007 

9:00 AM PST; SRDB Ref ID: 37532

Personal Communication with Health Physicist; Telephone Interview by ORAU Team; November 20, 

2007 11:00 AM PST; SRDB Ref ID: 37534 

Personal Communication with Health Physicist; Telephone Interview by ORAU Team; November 15, 

2007 10:30 AM PST; SRDB Ref ID: 37536 

Personal Communication with HP-Radiation Engineer; Telephone Interview by ORAU Team; 

November 29, 2007 1:00 PM PST; SRDB Ref ID: 37537 

Personal Communication with Health and Safety Manager; Telephone Interview by ORAU Team; 

November 30, 2007 9:00AM PST; SRDB Ref ID: 37538 

Petition Form B, 2007; Petition Form B; June 22, 2007; OSA Ref ID: 103252 

Reed, 1995, Dispute Surfaces on Rocketdyne Deaths Study, newspaper article; Mack Reed; July 11, 

1995; OSA Ref ID: 103253

Reed, 1997, Lab Linked to Cancer Risk, article in the Los Angeles Times; Mack Reed and Thomas H. 

Maugh; September 12, 1997; OSA Ref ID: 103253 

Rutherford, 2005, Sodium Reactor Experiment (SRE) Original Accident Release Data, memorandum 

to E. Crawford; Phil Rutherford; January 21, 2005; SRDB Ref ID: 19161 and 19162 

Rutherford, 2007, Potable Water, email correspondence between Eugene Potter and Philip 

Rutherford; Philip D. Rutherford; December 2007; SRDB Ref ID: 38782 

Rocketdyne, date unknown, The Rocket’s Red Glare, a short DVD from Rocketdynewatch.org 

SAIC, 1991, Interim Final RCRA Facility Assessment Report for Rockwell International Corporation, 

TZ4-R09015-RN-M07933; Science Applications International Corporation (SAIC), Technology 

Services Company; July 10, 1991; OSA Ref ID: 103715 

Sapere Consulting, 2005, Historical Site Assessment of Area IV Santa Susana Field Laboratory 

Ventura County, California, Volume 1-Methodology, DE-AC03-99SF21530; Sapere Consulting, Inc. 

and the Boeing Company; May 2005; SRDB Ref ID: 20531 

Sheperd, 1959, Quotation on Urinalyses, letter to Atomics International; J. L. Sheperd, May 19, 1959; 

SRDB Ref ID: 19056, pages 6-7. 
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Shleien, 1992, The Health Physics and Radiological Health Handbook, Revised Edition; Bernard 

Shleien-editor; Scinta, Inc., Silver Spring, MD; 1992; Publicly Available Document. 

SSFL, 2005, Employment History/Monitoring Record; Santa Susana Field Laboratory (SSFL); 

October 28, 2005; OSA Ref ID: 103253

Stewart, 1967, Commendation Regarding Uranium Fire, Canoga Facility, May 17, 1967, J. E. 

Stewart; May 19, 1967; OSA Ref ID: 103253 

United States District Court, 2007, Order Granting Plaintiffs’ Motion For Summary Judgment, 

Natural Resources Defense Council, Inc., Committee to Bridge the Gap, and City of Los Angeles 

(Plaintiffs) versus Department of Energy, Spencer Abraham, Secretary, Department of Energy, and 

Camile Yuan-Soo Hoo, Manager, National Nuclear Security Administration, Oakland Operations 

(Defendants); United States District Court; May 2, 2007; OSA Ref ID: 103253 

Excerpts from Unknown Report; unknown author; unknown date; OSA Ref ID: 103253, pages 151 

and 154 

Unknown author, unknown date, Project History Excerpts from Unknown Report; unknown author; 

unknown date; OSA Ref ID: 103253, pages 155 and 156 

Unknown author, 1958, Recommended Bioassay Procedures for Uranium; unknown author; October 

1958; SRDB Ref ID: 19672, page 7 

Unknown author, 1966, Radiation Safety Progress for November 1966, regarding a sodium reactor 

experiment; author unknown; December 9, 1966; OSA Ref ID: 103701

Unknown author, 1989, Various Pages from Journal, pictures and pages from article that was 

published in Ecology and Environment, Inc.; author not identified; July 19, 1989; OSA Ref ID: 

103253 

Various authors, 1960-1966, Collection of Routine Urine Specimen Collection Documents, Special 

Bioassay Request, and Routine Bioassay Specimen Requests; various authors; 1960-1966; SRDB Ref 

ID: 19053 

Various authors, 1970-1987, Various External Dosimetry Procedures at Area IV of the SSFL; 1970-

1987; SRDB Ref ID: 19143. 

Yoder, 2005, Landauer Film Information, email to Leo Faust; R. Craig Yoder; December 8, 2005; 

SRDB Ref ID: 19707